Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

Current and anticipated uses of thermal-hydraulic codes in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strizhov, V.; Kanukova, V.; Vinogradova, T.

1996-09-01

This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer frommore » melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.« less

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

DOE PAGES

Guo, Z.; Zweibaum, N.; Shao, M.; ...

2016-04-19

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate amore » staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. In conclusion, this paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.« less

Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

NASA Astrophysics Data System (ADS)

Sabotinov, Luben; Chevrier, Patrick

The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

Methodology, status and plans for development and assessment of the code ATHLET

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Austregesilo, H.; Lerchl, G.

1997-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The codemore » has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities.« less

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Thermal-hydraulic modeling needs for passive reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, J.M.

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered,more » but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.« less

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Current and anticipated uses of the thermal hydraulics codes at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caruso, R.

1997-07-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support thesemore » needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.« less

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

NASA Astrophysics Data System (ADS)

Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

2016-02-01

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level.

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guo, Z.; Zweibaum, N.; Shao, M.

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate amore » staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. In conclusion, this paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.« less

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Coupled field effects in BWR stability simulations using SIMULATE-3K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

Development of high-fidelity multiphysics system for light water reactor analysis

NASA Astrophysics Data System (ADS)

Magedanz, Jeffrey W.

There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining codes into a single executable, they are usually still developed and maintained separately. It should thus be a design objective to minimize the changes to those codes, and keep the changes to each code free of dependence on the details of the other codes. This will ease the incorporation of new versions of the code into the coupling, as well as re-use of parts of the coupling to couple with different codes. In order to fulfill this objective, an interface for each code was created in the form of an object-oriented abstract data type. Object-oriented programming is an effective method for enforcing a separation between different parts of a program, and clarifying the communication between them. The interfaces enable the main program to control the codes in terms of high-level functionality. This differs from the established practice of a master/slave relationship, in which the slave code is incorporated into the master code as a set of subroutines. While this PhD research continues previous work with a coupling between CTF and TORT-TD, it makes two major original contributions: (1) using a fuel-performance code, instead of a thermal-hydraulics code's simplified built-in models, to model the feedback from the fuel rods, and (2) the design of an object-oriented interface as an innovative method to interact with a coupled code in a high-level, easily-understandable manner. The resulting code system will serve as a tool to study the question of under what conditions, and to what extent, these higher-fidelity methods will provide benefits to reactor core analysis. (Abstract shortened by UMI.)

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

RETRANO3 benchmarks for Beaver Valley plant transients and FSAR analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beaumont, E.T.; Feltus, M.A.

1993-01-01

Any best-estimate code (e.g., RETRANO3) results must be validated against plant data and final safety analysis report (FSAR) predictions. The need for two independent means of benchmarking is necessary to ensure that the results were not biased toward a particular data set and to have a certain degree of accuracy. The code results need to be compared with previous results and show improvements over previous code results. Ideally, the two best means of benchmarking a thermal hydraulics code are comparing results from previous versions of the same code along with actual plant data. This paper describes RETRAN03 benchmarks against RETRAN02more » results, actual plant data, and FSAR predictions. RETRAN03, the Electric Power Research Institute's latest version of the RETRAN thermal-hydraulic analysis codes, offers several upgrades over its predecessor, RETRAN02 Mod5. RETRAN03 can use either implicit or semi-implicit numerics, whereas RETRAN02 Mod5 uses only semi-implicit numerics. Another major upgrade deals with slip model options. RETRAN03 added several new models, including a five-equation model for more accurate modeling of two-phase flow. RETPAN02 Mod5 should give similar but slightly more conservative results than RETRAN03 when executed with RETRAN02 Mod5 options.« less

Pump-stopping water hammer simulation based on RELAP5

NASA Astrophysics Data System (ADS)

Yi, W. S.; Jiang, J.; Li, D. D.; Lan, G.; Zhao, Z.

2013-12-01

RELAP5 was originally designed to analyze complex thermal-hydraulic interactions that occur during either postulated large or small loss-of-coolant accidents in PWRs. However, as development continued, the code was expanded to include many of the transient scenarios that might occur in thermal-hydraulic systems. The fast deceleration of the liquid results in high pressure surges, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increase. This phenomenon is called water hammer. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the system when the pressure surges become considerably high. If this happens and when the pressure exceeds the critical pressure that the pipe or the fittings along the pipeline can burden, it will result in the failure of the whole pipeline integrity. The purpose of this article is to introduce the RELAP5 to the simulation and analysis of water hammer situations. Based on the knowledge of the RELAP5 code manuals and some relative documents, the authors utilize RELAP5 to set up an example of water-supply system via an impeller pump to simulate the phenomena of the pump-stopping water hammer. By the simulation of the sample case and the subsequent analysis of the results that the code has provided, we can have a better understand of the knowledge of water hammer as well as the quality of the RELAP5 code when it's used in the water-hammer fields. In the meantime, By comparing the results of the RELAP5 based model with that of other fluid-transient analysis software say, PIPENET. The authors make some conclusions about the peculiarity of RELAP5 when transplanted into water-hammer research and offer several modelling tips when use the code to simulate a water-hammer related case.

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernnat, W.; Buck, M.; Mattes, M.

The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code ormore » memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)« less

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

FY17 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Jung, Y. S.; Smith, M. A.

2017-09-30

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H)more » capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.« less

Verification of the predictive capabilities of the 4C code cryogenic circuit model

NASA Astrophysics Data System (ADS)

Zanino, R.; Bonifetto, R.; Hoa, C.; Richard, L. Savoldi

2014-01-01

The 4C code was developed to model thermal-hydraulics in superconducting magnet systems and related cryogenic circuits. It consists of three coupled modules: a quasi-3D thermal-hydraulic model of the winding; a quasi-3D model of heat conduction in the magnet structures; an object-oriented a-causal model of the cryogenic circuit. In the last couple of years the code and its different modules have undergone a series of validation exercises against experimental data, including also data coming from the supercritical He loop HELIOS at CEA Grenoble. However, all this analysis work was done each time after the experiments had been performed. In this paper a first demonstration is given of the predictive capabilities of the 4C code cryogenic circuit module. To do that, a set of ad-hoc experimental scenarios have been designed, including different heating and control strategies. Simulations with the cryogenic circuit module of 4C have then been performed before the experiment. The comparison presented here between the code predictions and the results of the HELIOS measurements gives the first proof of the excellent predictive capability of the 4C code cryogenic circuit module.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavisca, M.; Ghaderi, M.; Khatib-Rahbar, M.

2006-07-01

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U.S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents themore » results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990's. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models. (authors)« less

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

A New Capability for Nuclear Thermal Propulsion Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

Performance of a parallel thermal-hydraulics code TEMPEST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fann, G.I.; Trent, D.S.

The authors describe the parallelization of the Tempest thermal-hydraulics code. The serial version of this code is used for production quality 3-D thermal-hydraulics simulations. Good speedup was obtained with a parallel diagonally preconditioned BiCGStab non-symmetric linear solver, using a spatial domain decomposition approach for the semi-iterative pressure-based and mass-conserved algorithm. The test case used here to illustrate the performance of the BiCGStab solver is a 3-D natural convection problem modeled using finite volume discretization in cylindrical coordinates. The BiCGStab solver replaced the LSOR-ADI method for solving the pressure equation in TEMPEST. BiCGStab also solves the coupled thermal energy equation. Scalingmore » performance of 3 problem sizes (221220 nodes, 358120 nodes, and 701220 nodes) are presented. These problems were run on 2 different parallel machines: IBM-SP and SGI PowerChallenge. The largest problem attains a speedup of 68 on an 128 processor IBM-SP. In real terms, this is over 34 times faster than the fastest serial production time using the LSOR-ADI solver.« less

Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

NASA Astrophysics Data System (ADS)

Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

VICTORIA: A mechanistic model for radionuclide behavior in the reactor coolant system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schaperow, J.H.; Bixler, N.E.

1996-12-31

VICTORIA is the U.S. Nuclear Regulatory Commission`s (NRC`s) mechanistic, best-estimate code for analysis of fission product release from the core and subsequent transport in the reactor vessel and reactor coolant system. VICTORIA requires thermal-hydraulic data (i.e., temperatures, pressures, and velocities) as input. In the past, these data have been taken from the results of calculations from thermal-hydraulic codes such as SCDAP/RELAP5, MELCOR, and MAAP. Validation and assessment of VICTORIA 1.0 have been completed. An independent peer review of VICTORIA, directed by Brookhaven National Laboratory and supported by experts in the areas of fuel release, fission product chemistry, and aerosol physics,more » has been undertaken. This peer review, which will independently assess the code`s capabilities, is nearing completion with the peer review committee`s final report expected in Dec 1996. A limited amount of additional development is expected as a result of the peer review. Following this additional development, the NRC plans to release VICTORIA 1.1 and an updated and improved code manual. Future plans mainly involve use of the code for plant calculations to investigate specific safety issues as they arise. Also, the code will continue to be used in support of the Phebus experiments.« less

Diffusive deposition of aerosols in Phebus containment during FPT-2 test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontautas, A.; Urbonavicius, E.

2012-07-01

At present the lumped-parameter codes is the main tool to investigate the complex response of the containment of Nuclear Power Plant in case of an accident. Continuous development and validation of the codes is required to perform realistic investigation of the processes that determine the possible source term of radioactive products to the environment. Validation of the codes is based on the comparison of the calculated results with the measurements performed in experimental facilities. The most extensive experimental program to investigate fission product release from the molten fuel, transport through the cooling circuit and deposition in the containment is performedmore » in PHEBUS test facility. Test FPT-2 performed in this facility is considered for analysis of processes taking place in containment. Earlier performed investigations using COCOSYS code showed that the code could be successfully used for analysis of thermal-hydraulic processes and deposition of aerosols, but there was also noticed that diffusive deposition on the vertical walls does not fit well with the measured results. In the CPA module of ASTEC code there is implemented different model for diffusive deposition, therefore the PHEBUS containment model was transferred from COCOSYS code to ASTEC-CPA to investigate the influence of the diffusive deposition modelling. Analysis was performed using PHEBUS containment model of 16 nodes. The calculated thermal-hydraulic parameters are in good agreement with measured results, which gives basis for realistic simulation of aerosol transport and deposition processes. Performed investigations showed that diffusive deposition model has influence on the aerosol deposition distribution on different surfaces in the test facility. (authors)« less

Advances in modelling of condensation phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas

2007-07-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code systemmore » ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)« less

The kinetics of aerosol particle formation and removal in NPP severe accidents

NASA Astrophysics Data System (ADS)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.; Dolganov, Rostislav A.

2016-06-01

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal-hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into the KUPOL-M thermal-hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1981-09-01

This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity.more » The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.« less

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE PAGES

Hu, Rui

2016-11-19

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

Numerical analysis of one-dimensional temperature data for groundwater/surface-water exchange with 1DTempPro

NASA Astrophysics Data System (ADS)

Voytek, E. B.; Drenkelfuss, A.; Day-Lewis, F. D.; Healy, R. W.; Lane, J. W.; Werkema, D. D.

2012-12-01

Temperature is a naturally occurring tracer, which can be exploited to infer the movement of water through the vadose and saturated zones, as well as the exchange of water between aquifers and surface-water bodies, such as estuaries, lakes, and streams. One-dimensional (1D) vertical temperature profiles commonly show thermal amplitude attenuation and increasing phase lag of diurnal or seasonal temperature variations with propagation into the subsurface. This behavior is described by the heat-transport equation (i.e., the convection-conduction-dispersion equation), which can be solved analytically in 1D under certain simplifying assumptions (e.g., sinusoidal or steady-state boundary conditions and homogeneous hydraulic and thermal properties). Analysis of 1D temperature profiles using analytical models provides estimates of vertical groundwater/surface-water exchange. The utility of these estimates can be diminished when the model assumptions are violated, as is common in field applications. Alternatively, analysis of 1D temperature profiles using numerical models allows for consideration of more complex and realistic boundary conditions. However, such analyses commonly require model calibration and the development of input files for finite-difference or finite-element codes. To address the calibration and input file requirements, a new computer program, 1DTempPro, is presented that facilitates numerical analysis of vertical 1D temperature profiles. 1DTempPro is a graphical user interface (GUI) to the USGS code VS2DH, which numerically solves the flow- and heat-transport equations. Pre- and post-processor features within 1DTempPro allow the user to calibrate VS2DH models to estimate groundwater/surface-water exchange and hydraulic conductivity in cases where hydraulic head is known. This approach improves groundwater/ surface-water exchange-rate estimates for real-world data with complexities ill-suited for examination with analytical methods. Additionally, the code allows for time-varying temperature and hydraulic boundary conditions. Here, we present the approach and include examples for several datasets from stream/aquifer systems.

Transient analysis of ”2 inch Direct Vessel Injection line break” in SPES-2 facility by using TRACE code

NASA Astrophysics Data System (ADS)

D'Amico, S.; Lombardo, C.; Moscato, I.; Polidori, M.; Vella, G.

2015-11-01

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis here presented was performed by using TRACE and CATHARE thermal-hydraulic codes with the purpose of evaluating their prediction capability. The main results show that the TRACE model well predicts the overall behaviour of the plant during the transient, in particular it is able to simulate the principal thermal-hydraulic phenomena related to all passive safety systems. The performance of the presented CATHARE noding has suggested some possible improvements of the model.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C

The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical modelmore » in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.« less

Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook

The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developedmore » a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT.« less

Thermal hydraulic simulations, error estimation and parameter sensitivity studies in Drekar::CFD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Thomas Michael; Shadid, John N.; Pawlowski, Roger P.

2014-01-01

This report describes work directed towards completion of the Thermal Hydraulics Methods (THM) CFD Level 3 Milestone THM.CFD.P7.05 for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Nuclear Hub effort. The focus of this milestone was to demonstrate the thermal hydraulics and adjoint based error estimation and parameter sensitivity capabilities in the CFD code called Drekar::CFD. This milestone builds upon the capabilities demonstrated in three earlier milestones; THM.CFD.P4.02 [12], completed March, 31, 2012, THM.CFD.P5.01 [15] completed June 30, 2012 and THM.CFD.P5.01 [11] completed on October 31, 2012.

75 FR 69140 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-10

... To Support Specific Success Criteria in the Standardized Plant Analysis Risk Models--Surry and Peach... INFORMATION: NUREG-1953, ``Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the... document entitled: NUREG-1953, ``Confirmatory Thermal- Hydraulic Analysis to Support Specific Success...

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Specifications for a coupled neutronics thermal-hydraulics SFR test case

NASA Astrophysics Data System (ADS)

Tassone, A.; Smirnov, A. D.; Tikhomirov, G. V.

2017-01-01

Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors are a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q‴), density and temperature for the fuel, the cladding and the coolant.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integratedmore » into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.« less

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

RAMONA-3B application to Browns Ferry ATWS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slovik, G.C.; Neymotin, L.; Cazzoli, E.

1984-01-01

This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve,more » and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.« less

Posttest calculations of bundle quench test CORA-13 with ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestele, J.; Trambauer, K.; Schubert, J.D.

Gesellschaft fuer Anlagen- und Reaktorsicherheit is developing, in cooperation with the Institut fuer Kernenergetik und Energiesysteme, Stuttgart, the system code Analysis of Thermalhydraulics of Leaks and Transients with Core Degradation (ATHLET-CD). The code consists of detailed models of the thermal hydraulics of the reactor coolant system. This thermo-fluid dynamics module is coupled with modules describing the early phase of the core degradation, like cladding deformation, oxidation and melt relocation, and the release and transport of fission products. The assessment of the code is being done by the analysis of separate effect tests, integral tests, and plant events. The code willmore » be applied to the verification of severe accident management procedures. The out-of-pile test CORA-13 was conducted by Forschungszentrum Karlsruhe in their CORA test facility. The test consisted of two phases, a heatup phase and a quench phase. At the beginning of the quench phase, a sharp peak in the hydrogen generation rate was observed. Both phases of the test have been calculated with the system code ATHLET-CD. Special efforts have been made to simulate the heat losses and the flow distribution in the test facility and the thermal hydraulics during the quench phase. In addition to previous calculations, the material relocation and the quench phase have been modeled. The temperature increase during the heatup phase, the starting time of the temperature escalation, and the maximum temperatures have been calculated correctly. At the beginning of the quench phase, an increased hydrogen generation rate has been calculated as measured in the experiment.« less

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

VERA and VERA-EDU 3.5 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

Thermal-hydraulic posttest analysis for the ANL/MCTF 360/sup 0/ model heat-exchanger water test under mixed convection. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, C.I.; Sha, W.T.; Kasza, K.E.

As a result of the uncertainties in the understanding of the influence of thermal-buoyancy effects on the flow and heat transfer in Liquid Metal Fast Breeder Reactor heat exchangers and steam generators under off-normal operating conditions, an extensive experimental program is being conducted at Argonne National Laboratory to eliminate these uncertainties. Concurrently, a parallel analytical effort is also being pursued to develop a three-dimensional transient computer code (COMMIX-IHX) to study and predict heat exchanger performance under mixed, forced, and free convection conditions. This paper presents computational results from a heat exchanger simulation and compares them with the results from amore » test case exhibiting strong thermal buoyancy effects. Favorable agreement between experiment and code prediction is obtained.« less

Assessment of the MHD capability in the ATHENA code using data from the ALEX facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roth, P.A.

1989-03-01

The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code is a system transient analysis code with multi-loop, multi-fluid capabilities, which is available to the fusion community at the National Magnetic Fusion Energy Computing Center (NMFECC). The work reported here assesses the ATHENA magnetohydrodynamic (MHD) pressure drop model for liquid metals flowing through a strong magnetic field. An ATHENA model was developed for two simple geometry, adiabatic test sections used in the Argonne Liquid Metal Experiment (ALEX) at Argonne National Laboratory (ANL). The pressure drops calculated by ATHENA agreed well with the experimental results from the ALEX facility.

Summary of papers on current and anticipated uses of thermal-hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caruso, R.

1997-07-01

The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especiallymore » faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Alfonsi; C. Rabiti; D. Mandelli

The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data miningmore » module« less

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Influence of geologic layering on heat transport and storage in an aquifer thermal energy storage system

NASA Astrophysics Data System (ADS)

Bridger, D. W.; Allen, D. M.

2014-01-01

A modeling study was carried out to evaluate the influence of aquifer heterogeneity, as represented by geologic layering, on heat transport and storage in an aquifer thermal energy storage (ATES) system in Agassiz, British Columbia, Canada. Two 3D heat transport models were developed and calibrated using the flow and heat transport code FEFLOW including: a "non-layered" model domain with homogeneous hydraulic and thermal properties; and, a "layered" model domain with variable hydraulic and thermal properties assigned to discrete geological units to represent aquifer heterogeneity. The base model (non-layered) shows limited sensitivity for the ranges of all thermal and hydraulic properties expected at the site; the model is most sensitive to vertical anisotropy and hydraulic gradient. Simulated and observed temperatures within the wells reflect a combination of screen placement and layering, with inconsistencies largely explained by the lateral continuity of high permeability layers represented in the model. Simulation of heat injection, storage and recovery show preferential transport along high permeability layers, resulting in longitudinal plume distortion, and overall higher short-term storage efficiencies.

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonablemore » accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.« less

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Austregesilo, H.; Velkov, K.

1997-07-01

The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.

Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

PubMed

Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

2016-12-01

Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

2014-06-01

Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

Benchmark Simulations of the Thermal-Hydraulic Responses during EBR-II Inherent Safety Tests using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui; Sumner, Tyler S.

2016-04-17

An advanced system analysis tool SAM is being developed for fast-running, improved-fidelity, and whole-plant transient analyses at Argonne National Laboratory under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. As an important part of code development, companion validation activities are being conducted to ensure the performance and validity of the SAM code. This paper presents the benchmark simulations of two EBR-II tests, SHRT-45R and BOP-302R, whose data are available through the support of DOE-NE’s Advanced Reactor Technology (ART) program. The code predictions of major primary coolant system parameter are compared with the test results. Additionally, the SAS4A/SASSYS-1 code simulationmore » results are also included for a code-to-code comparison.« less

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat @

2014-02-01

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Camous, F.; Jacq, F.; Chatelard, P.

1997-07-01

In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

Three-dimensional fuel pin model validation by prediction of hydrogen distribution in cladding and comparison with experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aly, A.; Avramova, Maria; Ivanov, Kostadin

To correctly describe and predict this hydrogen distribution there is a need for multi-physics coupling to provide accurate three-dimensional azimuthal, radial, and axial temperature distributions in the cladding. Coupled high-fidelity reactor-physics codes with a sub-channel code as well as with a computational fluid dynamics (CFD) tool have been used to calculate detailed temperature distributions. These high-fidelity coupled neutronics/thermal-hydraulics code systems are coupled further with the fuel-performance BISON code with a kernel (module) for hydrogen. Both hydrogen migration and precipitation/dissolution are included in the model. Results from this multi-physics analysis is validated utilizing calculations of hydrogen distribution using models informed bymore » data from hydrogen experiments and PIE data.« less

Sensitivity analysis of hydraulic and thermal parameters inducing anomalous heat flow in the Lower Yarmouk Gorge

NASA Astrophysics Data System (ADS)

Goretzki, Nora; Inbar, Nimrod; Kühn, Michael; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Siebert, Christian; Magri, Fabien

2016-04-01

The Lower Yarmouk Gorge, at the border between Israel and Jordan, is characterized by an anomalous temperature gradient of 46 °C/km. Numerical simulations of thermally-driven flow show that ascending thermal waters are the result of mixed convection, i.e. the interaction between the regional flow from the surrounding heights and buoyant flow within permeable faults [1]. Those models were calibrated against available temperature logs by running several forward problems (FP), with a classic "trial and error" method. In the present study, inverse problems (IP) are applied to find alternative parameter distributions that also lead to the observed thermal anomalies. The investigated physical parameters are hydraulic conductivity and thermal conductivity. To solve the IP, the PEST® code [2] is applied via the graphical interface FEPEST® in FEFLOW® [3]. The results show that both hydraulic and thermal conductivity are consistent with the values determined with the trial and error calibrations, which precede this study. However, the IP indicates that the hydraulic conductivity of the Senonian Paleocene aquitard can be 8.54*10-3 m/d, which is three times lower than the originally estimated value in [1]. Moreover, the IP suggests that the hydraulic conductivity in the faults can increase locally up to 0.17 m/d. These highly permeable areas can be interpreted as local damage zones at the faults/units intersections. They can act as lateral pathways in the deep aquifers that allow deep outflow of thermal water. This presentation provides an example about the application of FP and IP to infer a wide range of parameter values that reproduce observed environmental issues. [1] Magri F, Inbar N, Siebert C, Rosenthal E, Guttman J, Möller P (2015) Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520, 342-355 [2] Doherty J (2010) PEST: Model-Independent Parameter Estimation. user manual 5th Edition. Watermark, Brisbane, Australia [3] Diersch H.-J.G. (2014) FEFLOW Finite Element Modeling of Flow, Mass and Heat Transport in Porous and Fractured Media. Springer- Verlag Berlin Heidelberg, 996p

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

NASA Astrophysics Data System (ADS)

Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

2018-02-01

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies willmore » still be present in the successor code RELAP5/MOD3.« less

Assessment of the MHD capability in the ATHENA code using data from the ALEX (Argonne Liquid Metal Experiment) facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roth, P.A.

1988-10-28

The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code is a system transient analysis code with multi-loop, multi-fluid capabilities, which is available to the fusion community at the National Magnetic Fusion Energy Computing Center (NMFECC). The work reported here assesses the ATHENA magnetohydrodynamic (MHD) pressure drop model for liquid metals flowing through a strong magnetic field. An ATHENA model was developed for two simple geometry, adiabatic test sections used in the Argonne Liquid Metal Experiment (ALEX) at Argonne National Laboratory (ANL). The pressure drops calculated by ATHENA agreed well with the experimental results from the ALEX facility. 13 refs., 4more » figs., 2 tabs.« less

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

NASA Astrophysics Data System (ADS)

Katsuyama, Jinya; Uno, Shumpei; Watanabe, Tadashi; Li, Yinsheng

2018-03-01

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric Richard; Durbin, Samuel G

2007-04-01

The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

Thermal and flow analysis of the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% effort of Title 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steinke, R.G.; Mueller, C.; Knight, T.D.

1998-03-01

The computational fluid dynamics code CFX4.2 was used to evaluate steady-state thermal-hydraulic conditions in the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% of Title 1). Thirteen facility cases were evaluated with varying temperature dependence, drywell-array heat-source magnitude and distribution, location of the inlet tower, and no-flow curtains in the drywell-array vault. Four cases of a detailed model of the inlet-tower top fixture were evaluated to show the effect of the canopy-cruciform fixture design on the air pressure and flow distributions.

75 FR 80544 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-22

... To Support Specific Success Criteria in the Standardized Plant Analysis Risk Models--Surry and Peach... Specific Success Criteria in the Standardized Plant Analysis Risk Models--Surry and Peach Bottom, Draft..., ``Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis...

New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

2012-10-01

PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICSmore » (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.« less

Current and anticipated uses of thermal-hydraulic codes in Spain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pelayo, F.; Reventos, F.

1997-07-01

Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to themore » different uses of the codes, the applications mainly come from the authority, industry, universities and research institutions. The main conclusion derived from this paper establishes that further code development is justified if the following requisites are considered: (1) Safety relevance of scenarios not presently covered is established. (2) A substantial gain in margins or the capability to use realistic assumptions is obtained. (3) A general consensus on the licensability and methodology for application is reached. The role of Regulatory Body is stressed, as the most relevant outcome of the project may be related to the evolution of the licensing frame.« less

Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, P. T.; Dickson, T. L.; Yin, S.

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include themore » NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.« less

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

Portable Life Support Subsystem Thermal Hydraulic Performance Analysis

NASA Technical Reports Server (NTRS)

Barnes, Bruce; Pinckney, John; Conger, Bruce

2010-01-01

This paper presents the current state of the thermal hydraulic modeling efforts being conducted for the Constellation Space Suit Element (CSSE) Portable Life Support Subsystem (PLSS). The goal of these efforts is to provide realistic simulations of the PLSS under various modes of operation. The PLSS thermal hydraulic model simulates the thermal, pressure, flow characteristics, and human thermal comfort related to the PLSS performance. This paper presents modeling approaches and assumptions as well as component model descriptions. Results from the models are presented that show PLSS operations at steady-state and transient conditions. Finally, conclusions and recommendations are offered that summarize results, identify PLSS design weaknesses uncovered during review of the analysis results, and propose areas for improvement to increase model fidelity and accuracy.

Comparison of COBRA III-C and SABRE-1 (wire-wrap version) computational results with steady-state data from a 19-pin internally guard heated sodium-cooled bundle with a six-channel central blockage (THORS Bundle 3C). [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J F; Nelson, W R; Rose, S D

Computational thermal-hydraulic models of a 19-pin, electrically heated, wire-wrap liquid-metal fast breeder reactor test bundle were developed using two well-known subchannel analysis codes, COBRA III-C and SABRE-1 (wire-wrap version). These two codes use similar subchannel control volumes for the finite difference conservation equations but vary markedly in solution strategy and modeling capability. In particular, the empirical wire-wrap-forced diversion crossflow models are different. Surprisingly, however, crossflow velocity predictions of the two codes are very similar. Both codes show generally good agreement with experimental temperature data from a test in which a large radial temperature gradient was imposed. Differences between data andmore » code results are probably caused by experimental pin bowing, which is presently the limiting factor in validating coded empirical models.« less

A Method to Estimate the Hydraulic Conductivity of the Ground by TRT Analysis.

PubMed

Liuzzo Scorpo, Alberto; Nordell, Bo; Gehlin, Signhild

2017-01-01

The knowledge of hydraulic properties of aquifers is important in many engineering applications. Careful design of ground-coupled heat exchangers requires that the hydraulic characteristics and thermal properties of the aquifer must be well understood. Knowledge of groundwater flow rate and aquifer thermal properties is the basis for proper design of such plants. Different methods have been developed in order to estimate hydraulic conductivity by evaluating the transport of various tracers (chemical, heat etc.); thermal response testing (TRT) is a specific type of heat tracer that allows including the hydraulic properties in an effective thermal conductivity value. Starting from these considerations, an expeditious, graphical method was proposed to estimate the hydraulic conductivity of the aquifer, using TRT data and plausible assumption. Suggested method, which is not yet verified or proven to be reliable, should be encouraging further studies and development in this direction. © 2016, National Ground Water Association.

Thermal APU/hydraulics analysis program. User's guide and programmer's manual

NASA Technical Reports Server (NTRS)

Deluna, T. A.

1976-01-01

The User's Guide information plus program description necessary to run and have a general understanding of the Thermal APU/Hydraulics Analysis Program (TAHAP) is described. This information consists of general descriptions of the APU/hydraulic system and the TAHAP model, input and output data descriptions, and specific subroutine requirements. Deck setups and input data formats are included and other necessary and/or helpful information for using TAHAP is given. The math model descriptions for the driver program and each of its supporting subroutines are outlined.

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis - Midloop Operation with Loss of Residual Heat Removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulatedmore » by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G; Powers, Jeffrey J; Clarno, Kevin T

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this papermore » focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role ofmore » expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.« less

Process management using component thermal-hydraulic function classes

DOEpatents

Morman, James A.; Wei, Thomas Y. C.; Reifman, Jaques

1999-01-01

A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced.

Process management using component thermal-hydraulic function classes

DOEpatents

Morman, J.A.; Wei, T.Y.C.; Reifman, J.

1999-07-27

A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced. 5 figs.

A Comprehensive Validation Approach Using The RAVEN Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J

2015-06-01

The RAVEN computer code , developed at the Idaho National Laboratory, is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. RAVEN is a multi-purpose probabilistic and uncertainty quantification platform, capable to communicate with any system code. A natural extension of the RAVEN capabilities is the imple- mentation of an integrated validation methodology, involving several different metrics, that represent an evolution of the methods currently used in the field. The state-of-art vali- dation approaches use neither exploration of the input space through sampling strategies, nor a comprehensive variety of metrics neededmore » to interpret the code responses, with respect experimental data. The RAVEN code allows to address both these lacks. In the following sections, the employed methodology, and its application to the newer developed thermal-hydraulic code RELAP-7, is reported.The validation approach has been applied on an integral effect experiment, representing natu- ral circulation, based on the activities performed by EG&G Idaho. Four different experiment configurations have been considered and nodalized.« less

Systematic void fraction studies with RELAP5, FRANCESCA and HECHAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stosic, Z.; Preusche, G.

1996-08-01

In enhancing the scope of standard thermal-hydraulic codes applications beyond its capabilities, i.e. coupling with a one and/or three-dimensional kinetics core model, the void fraction, transferred from thermal-hydraulics to the core model, plays a determining role in normal operating range and high core flow, as the generated heat and axial power profiles are direct functions of void distribution in the core. Hence, it is very important to know if the void quality models in the programs which have to be coupled are compatible to allow the interactive exchange of data which are based on these constitutive void-quality relations. The presentedmore » void fraction study is performed in order to give the basis for the conclusion whether a transient core simulation using the RELAP5 void fractions can calculate the axial power shapes adequately. Because of that, the void fractions calculated with RELAP5 are compared with those calculated by BWR safety code for licensing--FRANCESCA and the best estimate model for pre- and post-dryout calculation in BWR heated channel--HECHAN. In addition, a comparison with standard experimental void-quality benchmark tube data is performed for the HECHAN code.« less

Conductor analysis of the ITER FEAT poloidal field coils during a plasma scenario

NASA Astrophysics Data System (ADS)

Nicollet, S.; Hertout, P.; Duchateau, J. L.; Bleyer, A.; Bessette, D.

2002-05-01

In the framework of the ITER (International Thermonuclear Experimental Reactor) FEAT (Fusion Energy Advanced Tokamak) project, a fully superconducting PF (Poloidal Field) system has been designed in detail. The Central Solenoid and the 6 equilibrium coils constituting the PF system provide the magnetic fields which develop, shape and control the 15 MA plasma during the 1800 s of a typical plasma scenario. The 6 PF coils will be wound two-in-hand from a 45 kA niobium-titanium CICC (Cable-In-Conduit-Conductor). These coils will experience severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation and AC losses (30 to 300 kJ). The AC losses along the PF coil pancakes are deduced from accurate magnetic field computations performed with a 3D magnetostatic code, TRAPS. The nuclear heating and the thermal radiation are assumed to be uniform over a given face of the PF coils. These heat loads are used as input to perform the thermal and hydraulic analysis with a finite element code, GANDALF. The temperature increases (0.1 to 0.4 K) are computed, the margins and performances of the conductor are evaluated.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bovalini, R.; D`Auria, F.; Galassi, G.M.

1996-11-01

RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool ofmore » ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions.« less

Improvements and applications of COBRA-TF for stand-alone and coupled LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, M.; Cuervo, D.; Ivanov, K.

2006-07-01

The advanced thermal-hydraulic subchannel code COBRA-TF has been recently improved and applied for stand-alone and coupled LWR core calculations at the Pennsylvania State Univ. in cooperation with AREVA NP GmbH (Germany)) and the Technical Univ. of Madrid. To enable COBRA-TF for academic and industrial applications including safety margins evaluations and LWR core design analyses, the code programming, numerics, and basic models were revised and substantially improved. The code has undergone through an extensive validation, verification, and qualification program. (authors)

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

NASA Astrophysics Data System (ADS)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

Constitutive relations in TRAC-P1A

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Saha, P.

1980-08-01

The purpose of this document is to describe the basic thermal-hydraulic models and correlations that are in the TRAC-P1A code, as released in March 1979. It is divided into two parts, A and B. Part A describes the models in the three-dimensional vessel module of TRAC, whereas Part B focuses on the loop components that are treated by one-dimensional formulations. The report follows the format of the questions prepared by the Analysis Development Branch of USNRC and the questionnaire has been attached to this document for completeness. Concerted efforts have been made in understanding the present models in TRAC-P1A bymore » going through the FORTRAN listing of the code. Some discrepancies between the code and the TRAC-P1A manual have been found. These are pointed out in this document. Efforts have also been made to check the TRAC references for the range of applicability of the models and correlations used in the code. 26 refs., 5 figs., 1 tab.« less

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code thatmore » is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.

1975-11-01

A mathematical model and a numerical solution scheme for thermal- hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media. (auth)

A User’s Guide to the PLTEMP/ANL Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Kalimullah, M.; Feldman, E. E.

2016-07-25

PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each withmore » its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Coupled calculation of the radiological release and the thermal-hydraulic behavior of a 3-loop PWR after a SGTR by means of the code RELAP5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Hove, W.; Van Laeken, K.; Bartsoen, L.

1995-09-01

To enable a more realistic and accurate calculation of the radiological consequences of a SGTR, a fission product transport model was developed. As the radiological releases strongly depend on the thermal-hydraulic transient, the model was included in the RELAP5 input decks of the Belgian NPPs. This enables the coupled calculation of the thermal-hydraulic transient and the radiological release. The fission product transport model tracks the concentration of the fission products in the primary circuit, in each of the SGs as well as in the condenser. This leads to a system of 6 coupled, first order ordinary differential equations with timemore » dependent coefficients. Flashing, scrubbing, atomisation and dry out of the break flow are accounted for. Coupling with the thermal-hydraulic calculation and correct modelling of the break position enables an accurate calculation of the mixture level above the break. Pre- and post-accident spiking in the primary circuit are introduced. The transport times in the FW-system and the SG blowdown system are also taken into account, as is the decontaminating effect of the primary make-up system and of the SG blowdown system. Physical input parameters such as the partition coefficients, half life times and spiking coefficients are explicitly introduced so that the same model can be used for iodine, caesium and noble gases.« less

PRATHAM: Parallel Thermal Hydraulics Simulations using Advanced Mesoscopic Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joshi, Abhijit S; Jain, Prashant K; Mudrich, Jaime A

2012-01-01

At the Oak Ridge National Laboratory, efforts are under way to develop a 3D, parallel LBM code called PRATHAM (PaRAllel Thermal Hydraulic simulations using Advanced Mesoscopic Methods) to demonstrate the accuracy and scalability of LBM for turbulent flow simulations in nuclear applications. The code has been developed using FORTRAN-90, and parallelized using the message passing interface MPI library. Silo library is used to compact and write the data files, and VisIt visualization software is used to post-process the simulation data in parallel. Both the single relaxation time (SRT) and multi relaxation time (MRT) LBM schemes have been implemented in PRATHAM.more » To capture turbulence without prohibitively increasing the grid resolution requirements, an LES approach [5] is adopted allowing large scale eddies to be numerically resolved while modeling the smaller (subgrid) eddies. In this work, a Smagorinsky model has been used, which modifies the fluid viscosity by an additional eddy viscosity depending on the magnitude of the rate-of-strain tensor. In LBM, this is achieved by locally varying the relaxation time of the fluid.« less

Flow and Temperature Distribution Evaluation on Sodium Heated Large-sized Straight Double-wall-tube Steam Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kisohara, Naoyuki; Moribe, Takeshi; Sakai, Takaaki

2006-07-01

The sodium heated steam generator (SG) being designed in the feasibility study on commercialized fast reactor cycle systems is a straight double-wall-tube type. The SG is large sized to reduce its manufacturing cost by economics of scale. This paper addresses the temperature and flow multi-dimensional distributions at steady state to obtain the prospect of the SG. Large-sized heat exchanger components are prone to have non-uniform flow and temperature distributions. These phenomena might lead to tube buckling or tube to tube-sheet junction failure in straight tube type SGs, owing to tubes thermal expansion difference. The flow adjustment devices installed in themore » SG are optimized to prevent these issues, and the temperature distribution properties are uncovered by analysis methods. The analysis model of the SG consists of two parts, a sodium inlet distribution plenum (the plenum) and a heat transfer tubes bundle region (the bundle). The flow and temperature distributions in the plenum and the bundle are evaluated by the three-dimensional code 'FLUENT' and the two dimensional thermal-hydraulic code 'MSG', respectively. The MSG code is particularly developed for sodium heated SGs in JAEA. These codes have revealed that the sodium flow is distributed uniformly by the flow adjustment devices, and that the lateral tube temperature distributions remain within the allowable temperature range for the structural integrity of the tubes and the tube to tube-sheet junctions. (authors)« less

CFD and Neutron codes coupling on a computational platform

NASA Astrophysics Data System (ADS)

Cerroni, D.; Da Vià, R.; Manservisi, S.; Menghini, F.; Scardovelli, R.

2017-01-01

In this work we investigate the thermal-hydraulics behavior of a PWR nuclear reactor core, evaluating the power generation distribution taking into account the local temperature field. The temperature field, evaluated using a self-developed CFD module, is exchanged with a neutron code, DONJON-DRAGON, which updates the macroscopic cross sections and evaluates the new neutron flux. From the updated neutron flux the new peak factor is evaluated and the new temperature field is computed. The exchange of data between the two codes is obtained thanks to their inclusion into the computational platform SALOME, an open-source tools developed by the collaborative project NURESAFE. The numerical libraries MEDmem, included into the SALOME platform, are used in this work, for the projection of computational fields from one problem to another. The two problems are driven by a common supervisor that can access to the computational fields of both systems, in every time step, the temperature field, is extracted from the CFD problem and set into the neutron problem. After this iteration the new power peak factor is projected back into the CFD problem and the new time step can be computed. Several computational examples, where both neutron and thermal-hydraulics quantities are parametrized, are finally reported in this work.

Heat Transfer Computations of Internal Duct Flows With Combined Hydraulic and Thermal Developing Length

NASA Technical Reports Server (NTRS)

Wang, C. R.; Towne, C. E.; Hippensteele, S. A.; Poinsatte, P. E.

1997-01-01

This study investigated the Navier-Stokes computations of the surface heat transfer coefficients of a transition duct flow. A transition duct from an axisymmetric cross section to a non-axisymmetric cross section, is usually used to connect the turbine exit to the nozzle. As the gas turbine inlet temperature increases, the transition duct is subjected to the high temperature at the gas turbine exit. The transition duct flow has combined development of hydraulic and thermal entry length. The design of the transition duct required accurate surface heat transfer coefficients. The Navier-Stokes computational method could be used to predict the surface heat transfer coefficients of a transition duct flow. The Proteus three-dimensional Navier-Stokes numerical computational code was used in this study. The code was first studied for the computations of the turbulent developing flow properties within a circular duct and a square duct. The code was then used to compute the turbulent flow properties of a transition duct flow. The computational results of the surface pressure, the skin friction factor, and the surface heat transfer coefficient were described and compared with their values obtained from theoretical analyses or experiments. The comparison showed that the Navier-Stokes computation could predict approximately the surface heat transfer coefficients of a transition duct flow.

BODYFIT-1FE: a computer code for three-dimensional steady-state/transient single-phase rod-bundle thermal-hydraulic analysis. Draft report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, B.C.J.; Sha, W.T.; Doria, M.L.

1980-11-01

The governing equations, i.e., conservation equations for mass, momentum, and energy, are solved as a boundary-value problem in space and an initial-value problem in time. BODYFIT-1FE code uses the technique of boundary-fitted coordinate systems where all the physical boundaries are transformed to be coincident with constant coordinate lines in the transformed space. By using this technique, one can prescribe boundary conditions accurately without interpolation. The transformed governing equations in terms of the boundary-fitted coordinates are then solved by using implicit cell-by-cell procedure with a choice of either central or upwind convective derivatives. It is a true benchmark rod-bundle code withoutmore » invoking any assumptions in the case of laminar flow. However, for turbulent flow, some empiricism must be employed due to the closure problem of turbulence modeling. The detailed velocity and temperature distributions calculated from the code can be used to benchmark and calibrate empirical coefficients employed in subchannel codes and porous-medium analyses.« less

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blyth, Taylor S.; Avramova, Maria

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR)more » cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.« less

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

NASA Astrophysics Data System (ADS)

Blyth, Taylor S.

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics-based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal-hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

Horizontal steam generator thermal-hydraulics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ubra, O.; Doubek, M.

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. Themore » 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.« less

Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Low, J.O.; Schmitt, B.E.

1988-02-01

A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may bemore » exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.« less

Numerical investigation of thermal-hydraulic performance of channel with protrusions by turbulent cross flow jet

NASA Astrophysics Data System (ADS)

Sahu, M. K.; Pandey, K. M.; Chatterjee, S.

2018-05-01

In this two dimensional numerical investigation, small rectangular channel with right angled triangular protrusions in the bottom wall of test section is considered. A slot nozzle is placed at the middle of top wall of channel which impinges air normal to the protruded surface. A duct flow and nozzle flow combined to form cross flow which is investigated for heat transfer enhancement of protruded channel. The governing equations for continuity, momentum, energy along with SST k-ω turbulence model are solved with finite volume based Computational fluid dynamics code ANSYS FLUENT 14.0. The range of duct Reynolds number considered for this analysis is 8357 to 51760. The ratios of pitch of protrusion to height of duct considered are 0.5, 0.64 and 0.82. The ratios of height of protrusion to height of duct considered are 0.14, 0.23 and 0.29. The effect of duct Reynolds number, pitch and height of protrusion on thermal-hydraulic performance is studied under cross flow condition. It is found that heat transfer rate is more at relatively larger pitch and small pressure drop is found in case of low height of protrusion.

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwon, Tae-Soon; Yun, Byong-Jo; Euh, Dong-Jin

Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focusedmore » on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.« less

MATCHED-INDEX-OF-REFRACTION FLOW FACILITY FOR FUNDAMENTAL AND APPLIED RESEARCH

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Carl Stoots; Donald M. McEligot

2014-11-01

Significant challenges face reactor designers with regard to thermal hydraulic design and associated modeling for advanced reactor concepts. Computational thermal hydraulic codes solve only a piece of the core. There is a need for a whole core dynamics system code with local resolution to investigate and understand flow behavior with all the relevant physics and thermo-mechanics. The matched index of refraction (MIR) flow facility at Idaho National Laboratory (INL) has a unique capability to contribute to the development of validated computational fluid dynamics (CFD) codes through the use of state-of-the-art optical measurement techniques, such as Laser Doppler Velocimetry (LDV) andmore » Particle Image Velocimetry (PIV). PIV is a non-intrusive velocity measurement technique that tracks flow by imaging the movement of small tracer particles within a fluid. At the heart of a PIV calculation is the cross correlation algorithm, which is used to estimate the displacement of particles in some small part of the image over the time span between two images. Generally, the displacement is indicated by the location of the largest peak. To quantify these measurements accurately, sophisticated processing algorithms correlate the locations of particles within the image to estimate the velocity (Ref. 1). Prior to use with reactor deign, the CFD codes have to be experimentally validated, which requires rigorous experimental measurements to produce high quality, multi-dimensional flow field data with error quantification methodologies. Computational thermal hydraulic codes solve only a piece of the core. There is a need for a whole core dynamics system code with local resolution to investigate and understand flow behavior with all the relevant physics and thermo-mechanics. Computational techniques with supporting test data may be needed to address the heat transfer from the fuel to the coolant during the transition from turbulent to laminar flow, including the possibility of an early laminarization of the flow (Refs. 2 and 3) (laminarization is caused when the coolant velocity is theoretically in the turbulent regime, but the heat transfer properties are indicative of the coolant velocity being in the laminar regime). Such studies are complicated enough that computational fluid dynamics (CFD) models may not converge to the same conclusion. Thus, experimentally scaled thermal hydraulic data with uncertainties should be developed to support modeling and simulation for verification and validation activities. The fluid/solid index of refraction matching technique allows optical access in and around geometries that would otherwise be impossible while the large test section of the INL system provides better spatial and temporal resolution than comparable facilities. Benchmark data for assessing computational fluid dynamics can be acquired for external flows, internal flows, and coupled internal/external flows for better understanding of physical phenomena of interest. The core objective of this study is to describe MIR and its capabilities, and mention current development areas for uncertainty quantification, mainly the uncertainty surface method and cross-correlation method. Using these methods, it is anticipated to establish a suitable approach to quantify PIV uncertainty for experiments performed in the MIR.« less

A hydrogen-oxygen rocket engine coolant passage design program (RECOP) for fluid-cooled thrust chambers and nozzles

NASA Technical Reports Server (NTRS)

Tomsik, Thomas M.

1994-01-01

The design of coolant passages in regeneratively cooled thrust chambers is critical to the operation and safety of a rocket engine system. Designing a coolant passage is a complex thermal and hydraulic problem requiring an accurate understanding of the heat transfer between the combustion gas and the coolant. Every major rocket engine company has invested in the development of thrust chamber computer design and analysis tools; two examples are Rocketdyne's REGEN code and Aerojet's ELES program. In an effort to augment current design capabilities for government and industry, the NASA Lewis Research Center is developing a computer model to design coolant passages for advanced regeneratively cooled thrust chambers. The RECOP code incorporates state-of-the-art correlations, numerical techniques and design methods, certainly minimum requirements for generating optimum designs of future space chemical engines. A preliminary version of the RECOP model was recently completed and code validation work is in progress. This paper introduces major features of RECOP and compares the analysis to design points for the first test case engine; the Pratt & Whitney RL10A-3-3A thrust chamber.

Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arndt, S.A.

1997-07-01

The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for codemore » use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Analysis of PANDA Passive Containment Cooling Steady-State Tests with the Spectra Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempniewicz, Marek M

2000-07-15

Results of post test simulation of the PANDA passive containment cooling (PCC) steady-state tests (S-series tests), performed at the PANDA facility at the Paul Scherrer Institute, Switzerland, are presented. The simulation has been performed using the computer code SPECTRA, a thermal-hydraulic code, designed specifically for analyzing containment behavior of nuclear power plants.Results of the present calculations are compared to the measurement data as well as the results obtained earlier with the codes MELCOR, TRAC-BF1, and TRACG. The calculated PCC efficiencies are somewhat lower than the measured values. Similar underestimation of PCC efficiencies had been obtained in the past, with themore » other computer codes. To explain this difference, it is postulated that condensate coming into the tubes forms a stream of liquid in one or two tubes, leaving most of the tubes unaffected. The condensate entering the water box is assumed to fall down in the form of droplets. With these assumptions, the results calculated with SPECTRA are close to the experimental data.It is concluded that the SPECTRA code is a suitable tool for analyzing containments of advanced reactors, equipped with passive containment cooling systems.« less

Analysis of steam generator tube rupture transients with single failure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

The Gesellschaft fuer Reaktorsicherheit is engaged in the collection and evaluation of light water reactor operating experience as well as analyses for the risk study of the pressurized water reactor (PWR). Within these activities, thermohydraulic calculations have been performed to show the influence of different boundary conditions and disturbances on the steam generator tube rupture (SGTR) transients. The analyses of these calculations have focused on the measures and systems needed to cope with an SGTR. The reference plant for this analysis is a 1300-MW(e) PWR of Kraftwerk Union design with four loops, each containing a U-tube steam generator (SG) andmore » a reactor cooling pump (RCP). The thermal-hydraulic code DRUFAN-02 was used for the transient calculations.« less

An analysis of the sliding pressure start-up of SCWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, F.; Yang, J.; Li, H.

In this paper, the preliminary sliding pressure start-up system and scheme of supercritical water-cooled reactor in CGNPC (CGN-SCWR) were proposed. Thermal-hydraulic behavior in start-up procedures was analyzed in detail by employing advanced reactor subchannel analysis software ATHAS. The maximum cladding temperature (MCT for short) and core power of fuel assembly during the whole start-up process were investigated comparatively. The results show that the recommended start-up scheme meets the design requirements from the perspective of thermal-hydraulic. (authors)

Study of premixing phase of steam explosion with JASMINE code in ALPHA program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu

Premixing phase of steam explosion has been studied in ALPHA Program at Japan Atomic Energy Research Institute (JAERI). An analytical model to simulate the premixing phase, JASMINE (JAERI Simulator for Multiphase Interaction and Explosion), has been developed based on a multi-dimensional multi-phase thermal hydraulics code MISTRAL (by Fuji Research Institute Co.). The original code was extended to simulate the physics in the premixing phenomena. The first stage of the code validation was performed by analyzing two mixing experiments with solid particles and water: the isothermal experiment by Gilbertson et al. (1992) and the hot particle experiment by Angelini et al.more » (1993) (MAGICO). The code predicted reasonably well the experiments. Effectiveness of the TVD scheme employed in the code was also demonstrated.« less

Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

DOE PAGES

Hu, Po; Wilson, Paul

2014-01-01

The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in themore » code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.« less

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason Hales; Various

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale codemore » that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Hydraulic manipulator design, analysis, and control at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kress, R.L.; Jansen, J.F.; Love, L.J.

1996-09-01

To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned to hydraulics as a means of actuation. Hydraulics have always been the actuator of choice when designing heavy-life construction and mining equipment such as bulldozers, backhoes, and tunneling devices. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem) sophisticated modeling, analysis, and control experiments are usually needed. To support the development and deployment of new hydraulic manipulators Oak Ridge National Laboratory (ORNL) has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators,more » hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The hydraulics laboratory at ORNL has three different manipulators. First is a 6-Degree-of-Freedom (6-DoF), multi-planer, teleoperated, flexible controls test bed used for the development of waste tank clean-up manipulator controls, thermal studies, system characterization, and manipulator tracking. Finally, is a human amplifier test bed used for the development of an entire new class of teleoperated systems. To compliment the hardware in the hydraulics laboratory, ORNL has developed a hydraulics simulation capability including a custom package to model the hydraulic systems and manipulators for performance studies and control development. This paper outlines the history of hydraulic manipulator developments at ORNL, describes the hydraulics laboratory, discusses the use of the equipment within the laboratory, and presents some of the initial results from experiments and modeling associated with these hydraulic manipulators. Included are some of the results from the development of the human amplifier/de-amplifier concepts, the characterization of the thermal sensitivity of hydraulic systems, and end-point tracking accuracy studies. Experimental and analytical results are included.« less

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meng Lin; Dong Hou; Zhihong Xu

2006-07-01

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, justmore » can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I and C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results. (authors)« less

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Numerical Simulation and Analyses of the Loss of Feedwater Transient at the Unit 4 of Kola NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevanovic, Vladimir D.; Stosic, Zoran V.; Kiera, Michael

2002-07-01

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give completemore » insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with safety codes. (authors)« less

Heat, Mass and Aerosol Transfers in Spray Conditions for Containment Application

NASA Astrophysics Data System (ADS)

Porcheron, Emmanuel; Lemaitre, Pascal; Nuboer, Amandine; Vendel, Jacques

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surété Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Inverse problem to constrain the controlling parameters of large-scale heat transport processes: The Tiberias Basin example

NASA Astrophysics Data System (ADS)

Goretzki, Nora; Inbar, Nimrod; Siebert, Christian; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Magri, Fabien

2015-04-01

Salty and thermal springs exist along the lakeshore of the Sea of Galilee, which covers most of the Tiberias Basin (TB) in the northern Jordan- Dead Sea Transform, Israel/Jordan. As it is the only freshwater reservoir of the entire area, it is important to study the salinisation processes that pollute the lake. Simulations of thermohaline flow along a 35 km NW-SE profile show that meteoric and relic brines are flushed by the regional flow from the surrounding heights and thermally induced groundwater flow within the faults (Magri et al., 2015). Several model runs with trial and error were necessary to calibrate the hydraulic conductivity of both faults and major aquifers in order to fit temperature logs and spring salinity. It turned out that the hydraulic conductivity of the faults ranges between 30 and 140 m/yr whereas the hydraulic conductivity of the Upper Cenomanian aquifer is as high as 200 m/yr. However, large-scale transport processes are also dependent on other physical parameters such as thermal conductivity, porosity and fluid thermal expansion coefficient, which are hardly known. Here, inverse problems (IP) are solved along the NW-SE profile to better constrain the physical parameters (a) hydraulic conductivity, (b) thermal conductivity and (c) thermal expansion coefficient. The PEST code (Doherty, 2010) is applied via the graphical interface FePEST in FEFLOW (Diersch, 2014). The results show that both thermal and hydraulic conductivity are consistent with the values determined with the trial and error calibrations. Besides being an automatic approach that speeds up the calibration process, the IP allows to cover a wide range of parameter values, providing additional solutions not found with the trial and error method. Our study shows that geothermal systems like TB are more comprehensively understood when inverse models are applied to constrain coupled fluid flow processes over large spatial scales. References Diersch, H.-J.G., 2014. FEFLOW Finite Element Modeling of Flow, Mass and Heat Transport in Porous and Fractured Media. Springer- Verlag Berlin Heidelberg ,996p. Doherty J., 2010, PEST: Model-Independent Parameter Estimation. user manual 5th Edition. Watermark, Brisbane, Australia Magri, F., Inbar, N., Siebert C., Rosenthal, E., Guttman, J., Möller, P., 2015. Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520(0), 342-355.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bakosi, Jozsef; Christon, Mark A.; Francois, Marianne M.

Progress is reported on computational capabilities for the grid-to-rod-fretting (GTRF) problem of pressurized water reactors. Numeca's Hexpress/Hybrid mesh generator is demonstrated as an excellent alternative to generating computational meshes for complex flow geometries, such as in GTRF. Mesh assessment is carried out using standard industrial computational fluid dynamics practices. Hydra-TH, a simulation code developed at LANL for reactor thermal-hydraulics, is demonstrated on hybrid meshes, containing different element types. A series of new Hydra-TH calculations has been carried out collecting turbulence statistics. Preliminary results on the newly generated meshes are discussed; full analysis will be documented in the L3 milestone, THM.CFD.P5.05,more » Sept. 2012.« less

Analysis of the SL-1 Accident Using RELAPS5-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francisco, A.D. and Tomlinson, E. T.

2007-11-08

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

Three-Dimensional Modeling of Fracture Clusters in Geothermal Reservoirs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghassemi, Ahmad

The objective of this is to develop a 3-D numerical model for simulating mode I, II, and III (tensile, shear, and out-of-plane) propagation of multiple fractures and fracture clusters to accurately predict geothermal reservoir stimulation using the virtual multi-dimensional internal bond (VMIB). Effective development of enhanced geothermal systems can significantly benefit from improved modeling of hydraulic fracturing. In geothermal reservoirs, where the temperature can reach or exceed 350oC, thermal and poro-mechanical processes play an important role in fracture initiation and propagation. In this project hydraulic fracturing of hot subsurface rock mass will be numerically modeled by extending the virtual multiplemore » internal bond theory and implementing it in a finite element code, WARP3D, a three-dimensional finite element code for solid mechanics. The new constitutive model along with the poro-thermoelastic computational algorithms will allow modeling the initiation and propagation of clusters of fractures, and extension of pre-existing fractures. The work will enable the industry to realistically model stimulation of geothermal reservoirs. The project addresses the Geothermal Technologies Office objective of accurately predicting geothermal reservoir stimulation (GTO technology priority item). The project goal will be attained by: (i) development of the VMIB method for application to 3D analysis of fracture clusters; (ii) development of poro- and thermoelastic material sub-routines for use in 3D finite element code WARP3D; (iii) implementation of VMIB and the new material routines in WARP3D to enable simulation of clusters of fractures while accounting for the effects of the pore pressure, thermal stress and inelastic deformation; (iv) simulation of 3D fracture propagation and coalescence and formation of clusters, and comparison with laboratory compression tests; and (v) application of the model to interpretation of injection experiments (planned by our industrial partner) with reference to the impact of the variations in injection rate and temperature, rock properties, and in-situ stress.« less

NGNP Data Management and Analysis System Modeling Capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cynthia D. Gentillon

2009-09-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the thirdmore » NDMAS objective. It describes capabilities for displaying the data in meaningful ways and identifying relationships among the measured quantities that contribute to their understanding.« less

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozlowski, T.; Downar, T.; Xu, Y.

2012-07-01

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validationmore » for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolstad, J.W.; Haarman, R.A.

The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid. Calculations show that continued injection of high pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island.

Joint three-dimensional inversion of coupled groundwater flow and heat transfer based on automatic differentiation: sensitivity calculation, verification, and synthetic examples

NASA Astrophysics Data System (ADS)

Rath, V.; Wolf, A.; Bücker, H. M.

2006-10-01

Inverse methods are useful tools not only for deriving estimates of unknown parameters of the subsurface, but also for appraisal of the thus obtained models. While not being neither the most general nor the most efficient methods, Bayesian inversion based on the calculation of the Jacobian of a given forward model can be used to evaluate many quantities useful in this process. The calculation of the Jacobian, however, is computationally expensive and, if done by divided differences, prone to truncation error. Here, automatic differentiation can be used to produce derivative code by source transformation of an existing forward model. We describe this process for a coupled fluid flow and heat transport finite difference code, which is used in a Bayesian inverse scheme to estimate thermal and hydraulic properties and boundary conditions form measured hydraulic potentials and temperatures. The resulting derivative code was validated by comparison to simple analytical solutions and divided differences. Synthetic examples from different flow regimes demonstrate the use of the inverse scheme, and its behaviour in different configurations.

1DTempPro V2: new features for inferring groundwater/surface-water exchange

USGS Publications Warehouse

Koch, Franklin W.; Voytek, Emily B.; Day-Lewis, Frederick D.; Healy, Richard W.; Briggs, Martin A.; Lane, John W.; Werkema, Dale D.

2016-01-01

A new version of the computer program 1DTempPro extends the original code to include new capabilities for (1) automated parameter estimation, (2) layer heterogeneity, and (3) time-varying specific discharge. The code serves as an interface to the U.S. Geological Survey model VS2DH and supports analysis of vertical one-dimensional temperature profiles under saturated flow conditions to assess groundwater/surface-water exchange and estimate hydraulic conductivity for cases where hydraulic head is known.

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cynthia D. Gentillon

2010-09-01

Projects for the Very High Temperature Reactor Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The Very High Temperature Reactor Technology Development Office has established the NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory to ensure that very high temperature reactor data are (1) qualified for use, (2) stored in amore » readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities.« less

Needs and opportunities for CFD-code validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, B.L.

1996-06-01

The conceptual design for the ESS target consists of a horizontal cylinder containing a liquid metal - mercury is considered in the present study - which circulates by forced convection and carries away the waste heat generated by the spallation reactions. The protons enter the target via a beam window, which must withstand the thermal, mechanical and radiation loads to which it is subjected. For a beam power of 5MW, it is estimated that about 3.3MW of waste heat would be deposited in the target material and associated structures. it is intended to confirm, by detailed thermal-hydraulics calculations, that amore » convective flow of the liquid metal target material can effectively remove the waste heat. The present series of Computational Fluid Dynamics (CFD) calculations has indicated that a single-inlet Target design leads to excessive local overheating, but a multiple-inlet design, is coolable. With this option, inlet flow streams, two from the sides and one from below, merge over the target window, cooling the window itself in crossflow and carrying away the heat generated volumetrically in the mercury with a strong axial flow down the exit channel. The three intersecting streams form a complex, three-dimensional, swirling flow field in which critical heat transfer processes are taking place. In order to produce trustworthy code simulations, it is necessary that the mesh resolution is adequate for the thermal-hydraulic conditions encountered and that the physical models used by the code are appropriate to the fluid dynamic environment. The former relies on considerable user experience in the application of the code, and the latter assurance is best gained in the context of controlled benchmark activities where measured data are available. Such activities will serve to quantify the accuracy of given models and to identify potential problem area for the numerical simulation which may not be obvious from global heat and mass balance considerations.« less

10. Floor Layout of Thermal Hydraulics Laboratory, from The Thermal ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Floor Layout of Thermal Hydraulics Laboratory, from The Thermal Hydraulics Laboratory at Hanford. General Electric Company, Hanford Atomic Products Operation, Richland, Washington, 1961. - D-Reactor Complex, Deaeration Plant-Refrigeration Buildings, Area 100-D, Richland, Benton County, WA

Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

2013-01-01

Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

NASA Astrophysics Data System (ADS)

Raffray, A. René; Federici, Gianfranco

1997-04-01

RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D’Auria, F; Rohatgi, Upendra S.

The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

NASA Astrophysics Data System (ADS)

Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

2015-09-01

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

The 25 kWe solar thermal Stirling hydraulic engine system: Conceptual design

NASA Technical Reports Server (NTRS)

White, Maurice; Emigh, Grant; Noble, Jack; Riggle, Peter; Sorenson, Torvald

1988-01-01

The conceptual design and analysis of a solar thermal free-piston Stirling hydraulic engine system designed to deliver 25 kWe when coupled to a 11 meter test bed concentrator is documented. A manufacturing cost assessment for 10,000 units per year was made. The design meets all program objectives including a 60,000 hr design life, dynamic balancing, fully automated control, more than 33.3 percent overall system efficiency, properly conditioned power, maximum utilization of annualized insolation, and projected production costs. The system incorporates a simple, rugged, reliable pool boiler reflux heat pipe to transfer heat from the solar receiver to the Stirling engine. The free-piston engine produces high pressure hydraulic flow which powers a commercial hydraulic motor that, in turn, drives a commercial rotary induction generator. The Stirling hydraulic engine uses hermetic bellows seals to separate helium working gas from hydraulic fluid which provides hydrodynamic lubrication to all moving parts. Maximum utilization of highly refined, field proven commercial components for electric power generation minimizes development cost and risk.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

NASA Astrophysics Data System (ADS)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) andmore » ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models.« less

CTF Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, Maria N.; Salko, Robert K.

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, andmore » subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.« less

Burner liner thermal-structural load modeling

NASA Technical Reports Server (NTRS)

Maffeo, R.

1986-01-01

The software package Transfer Analysis Code to Interface Thermal/Structural Problems (TRANCITS) was developed. The TRANCITS code is used to interface temperature data between thermal and structural analytical models. The use of this transfer module allows the heat transfer analyst to select the thermal mesh density and thermal analysis code best suited to solve the thermal problem and gives the same freedoms to the stress analyst, without the efficiency penalties associated with common meshes and the accuracy penalties associated with the manual transfer of thermal data.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness ofmore » the PIUS concept to severe off-normal conditions having a very low probability of occurrence.« less

Imaging hydraulic fractures using temperature transients in the Belridge Diatomite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shahin, G.T.; Johnston, R.M.

1995-12-31

Results of a temperature transient analysis of Shell`s Phase 1 and Phase 2 Diatomite Steamdrive Pilots are used to image hydraulic injection fracture lengths, angles, and heat injectivities into the low-permeability formation. The Phase 1 Pilot is a limited-interval injection test. In Phase 2, steam is injected into two 350 ft upper and lower zones through separate hydraulic fractures. Temperature response of both pilots is monitored with sixteen logging observation wells. A perturbation analysis of the non-linear pressure diffusion and heat transport equations indicates that at a permeability of about 0.1 md or less, heat transport in the Diatomite tendsmore » to be dominated by thermal diffusivity, and pressure diffusion is dominated by the ratio of thermal expansion to fluid compressibility. Under these conditions, the temperature observed at a logging observation well is governed by a dimensionless quantity that depends on the perpendicular distance between the observation well and the hydraulic fracture, divided by the square root of time. Using this dependence, a novel method is developed for imaging hydraulic fracture geometry and relative heat injectivity from the temperature history of the pilot.« less

RELAP5 Model of the First Wall/Blanket Primary Heat Transfer System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

2010-06-01

ITER inductive power operation is modeled and simulated using a system level computer code to evaluate the behavior of the Primary Heat Transfer System (PHTS) and predict parameter operational ranges. The control algorithm strategy and derivation are summarized in this report as well. A major feature of ITER is pulsed operation. The plasma does not burn continuously, but the power is pulsed with large periods of zero power between pulses. This feature requires active temperature control to maintain a constant blanket inlet temperature and requires accommodation of coolant thermal expansion during the pulse. In view of the transient nature ofmore » the power (plasma) operation state a transient system thermal-hydraulics code was selected: RELAP5. The code has a well-documented history for nuclear reactor transient analyses, it has been benchmarked against numerous experiments, and a large user database of commonly accepted modeling practices exists. The process of heat deposition and transfer in the blanket modules is multi-dimensional and cannot be accurately captured by a one-dimensional code such as RELAP5. To resolve this, a separate CFD calculation of blanket thermal power evolution was performed using the 3-D SC/Tetra thermofluid code. A 1D-3D co-simulation more realistically models FW/blanket internal time-dependent thermal inertia while eliminating uncertainties in the time constant assumed in a 1-D system code. Blanket water outlet temperature and heat release histories for any given ITER pulse operation scenario are calculated. These results provide the basis for developing time dependent power forcing functions which are used as input in the RELAP5 calculations.« less

Thermally Actuated Hydraulic Pumps

NASA Technical Reports Server (NTRS)

Jones, Jack; Ross, Ronald; Chao, Yi

2008-01-01

Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research vessels. Heretofore, electrically actuated hydraulic pumps have been used for this purpose. By eliminating the demand for electrical energy for pumping, the use of the thermally actuated hydraulic pumps could prolong the intervals between battery charges, thus making it possible to greatly increase the durations of undersea exploratory missions.

Thermal hydraulic behavior and efficiency analysis of an all-vanadium redox flow battery

NASA Astrophysics Data System (ADS)

Xiong, Binyu; Zhao, Jiyun; Tseng, K. J.; Skyllas-Kazacos, Maria; Lim, Tuti Mariana; Zhang, Yu

2013-11-01

Vanadium redox flow batteries (VRBs) are very competitive for large-capacity energy storage in power grids and in smart buildings due to low maintenance costs, high design flexibility, and long cycle life. Thermal hydraulic modeling of VRB energy storage systems is an important issue and temperature has remarkable impacts on the battery efficiency, the lifetime of material and the stability of the electrolytes. In this paper, a lumped model including auxiliary pump effect is developed to investigate the VRB temperature responses under different operating and surrounding environmental conditions. The impact of electrolyte flow rate and temperature on the battery electrical characteristics and efficiencies are also investigated. A one kilowatt VRB system is selected to conduct numerical simulations. The thermal hydraulic model is benchmarked with experimental data and good agreement is found. Simulation results show that pump power is sensitive to hydraulic design and flow rates. The temperature in the stack and tanks rises up about 10 °C under normal operating conditions for the stack design and electrolyte volume selected. An optimal flow rate of around 90 cm3 s-1 is obtained for the proposed battery configuration to maximize battery efficiency. The models developed in this paper can also be used for the development of a battery control strategy to achieve satisfactory thermal hydraulic performance and maximize energy efficiency.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Curtis; Mandelli, Diego; Prescott, Steven

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools.more » This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.« less

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paniagua, J.; Rohatgi, U.S.; Prasad, V.

1996-10-01

RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

The COSIMA experiments and their verification, a data base for the validation of two phase flow computer codes

NASA Astrophysics Data System (ADS)

Class, G.; Meyder, R.; Stratmanns, E.

1985-12-01

The large data base for validation and development of computer codes for two-phase flow, generated at the COSIMA facility, is reviewed. The aim of COSIMA is to simulate the hydraulic, thermal, and mechanical conditions in the subchannel and the cladding of fuel rods in pressurized water reactors during the blowout phase of a loss of coolant accident. In terms of fuel rod behavior, it is found that during blowout under realistic conditions only small strains are reached. For cladding rupture extremely high rod internal pressures are necessary. The behavior of fuel rod simulators and the effect of thermocouples attached to the cladding outer surface are clarified. Calculations performed with the codes RELAP and DRUFAN show satisfactory agreement with experiments. This can be improved by updating the phase separation models in the codes.

Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soli Khericha; Edwin Harvego; John Svoboda

2012-01-01

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstratemore » Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, S. M.; Kim, K. Y.

Printed circuit heat exchanger (PCHE) is recently considered as a recuperator for the high temperature gas cooled reactor. In this work, the zigzag-channels of a PCHE have been optimized by using three-dimensional Reynolds-Averaged Navier-Stokes (RANS) analysis and response surface approximation (RSA) modeling technique to enhance thermal-hydraulic performance. Shear stress transport turbulence model is used as a turbulence closure. The objective function is defined as a linear combination of the functions related to heat transfer and friction loss of the PCHE, respectively. Three geometric design variables viz., the ratio of the radius of the fillet to hydraulic diameter of the channels,more » the ratio of wavelength to hydraulic diameter of the channels, and the ratio of wave height to hydraulic diameter of the channels, are used for the optimization. Design points are selected through Latin-hypercube sampling. The optimal design is determined through the RSA model which uses RANS derived calculations at the design points. The results show that the optimum shape enhances considerably the thermal-hydraulic performance than a reference shape. (authors)« less

Hydraulic performance of compacted clay liners under simulated daily thermal cycles.

PubMed

Aldaeef, A A; Rayhani, M T

2015-10-01

Compacted clay liners (CCLs) are commonly used as hydraulic barriers in several landfill applications to isolate contaminants from the surrounding environment and minimize the escape of leachate from the landfill. Prior to waste placement in landfills, CCLs are often exposed to temperature fluctuations which can affect the hydraulic performance of the liner. Experimental research was carried out to evaluate the effects of daily thermal cycles on the hydraulic performance of CCLs under simulated landfill conditions. Hydraulic conductivity tests were conducted on different soil specimens after being exposed to various thermal and dehydration cycles. An increase in the CCL hydraulic conductivity of up to one order of magnitude was recorded after 30 thermal cycles for soils with low plasticity index (PI = 9.5%). However, medium (PI = 25%) and high (PI = 37.2%) plasticity soils did not show significant hydraulic deviation due to their self-healing potential. Overlaying the CCL with a cover layer minimized the effects of daily thermal cycles, and maintained stable hydraulic performance in the CCLs even after exposure to 60 thermal cycles. Wet-dry cycles had a significant impact on the hydraulic aspect of low plasticity CCLs. However, medium and high plasticity CCLs maintained constant hydraulic performance throughout the test intervals. The study underscores the importance of protecting the CCL from exposure to atmosphere through covering it by a layer of geomembrane or an interim soil layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

A Novel Multi-Scale Domain Overlapping CFD/STH Coupling Methodology for Multi-Dimensional Flows Relevant to Nuclear Applications

NASA Astrophysics Data System (ADS)

Grunloh, Timothy P.

The objective of this dissertation is to develop a 3-D domain-overlapping coupling method that leverages the superior flow field resolution of the Computational Fluid Dynamics (CFD) code STAR-CCM+ and the fast execution of the System Thermal Hydraulic (STH) code TRACE to efficiently and accurately model thermal hydraulic transport properties in nuclear power plants under complex conditions of regulatory and economic importance. The primary contribution is the novel Stabilized Inertial Domain Overlapping (SIDO) coupling method, which allows for on-the-fly correction of TRACE solutions for local pressures and velocity profiles inside multi-dimensional regions based on the results of the CFD simulation. The method is found to outperform the more frequently-used domain decomposition coupling methods. An STH code such as TRACE is designed to simulate large, diverse component networks, requiring simplifications to the fluid flow equations for reasonable execution times. Empirical correlations are therefore required for many sub-grid processes. The coarse grids used by TRACE diminish sensitivity to small scale geometric details such as Reactor Pressure Vessel (RPV) internals. A CFD code such as STAR-CCM+ uses much finer computational meshes that are sensitive to the geometric details of reactor internals. In turbulent flows, it is infeasible to fully resolve the flow solution, but the correlations used to model turbulence are at a low level. The CFD code can therefore resolve smaller scale flow processes. The development of a 3-D coupling method was carried out with the intention of improving predictive capabilities of transport properties in the downcomer and lower plenum regions of an RPV in reactor safety calculations. These regions are responsible for the multi-dimensional mixing effects that determine the distribution at the core inlet of quantities with reactivity implications, such as fluid temperature and dissolved neutron absorber concentration.

Multidimensional computer simulation of Stirling cycle engines

NASA Technical Reports Server (NTRS)

Hall, C. A.; Porsching, T. A.; Medley, J.; Tew, R. C.

1990-01-01

The computer code ALGAE (algorithms for the gas equations) treats incompressible, thermally expandable, or locally compressible flows in complicated two-dimensional flow regions. The solution method, finite differencing schemes, and basic modeling of the field equations in ALGAE are applicable to engineering design settings of the type found in Stirling cycle engines. The use of ALGAE to model multiple components of the space power research engine (SPRE) is reported. Videotape computer simulations of the transient behavior of the working gas (helium) in the heater-regenerator-cooler complex of the SPRE demonstrate the usefulness of such a program in providing information on thermal and hydraulic phenomena in multiple component sections of the SPRE.

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

77 FR 9707 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Thermal-Hydraulics...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Thermal-Hydraulics Phenomena; Revision to February 22, 2012, ACRS Meeting Federal Register Notice The Federal Register Notice for the ACRS Subcommittee meeting on Thermal-Hydraulics Phenomena...

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Coupling procedure for TRANSURANUS and KTF codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jimenez, J.; Alglave, S.; Avramova, M.

2012-07-01

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned toolsmore » can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)« less

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

PubMed

Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-08-06

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

PubMed Central

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-01-01

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

Preliminary Analysis of a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson, J. Boise

2006-01-01

A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A simple 1-D thermal model indicates the necessity of natural convection to maintain acceptable temperatures and pressures in the water shield. CFD analysis is done to quantify the natural convection in the shield, and predicts sufficient natural convection to transfer heat through the shield with small temperature gradients. A test program will he designed to experimentally verify the thermal hydraulic performance of the shield, and to anchor the CFD models to experimental results.

THE RETC CODE FOR QUANTIFYING THE HYDRAULIC FUNCTIONS OF UNSATURATED SOILS

EPA Science Inventory

This report describes the RETC computer code for analyzing the soil water retention and hydraulic conductivity functions of unsaturated soils. These hydraulic properties are key parameters in any quantitative description of water flow into and through the unsaturated zone of soil...

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Nathan C.; Gauntt, Randall O.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of themore » of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.« less

Project Summary. THE RETC CODE FOR QUANTIFYING THE HYDRAULIC FUNCTIONS OF UNSATURATED SOILS

EPA Science Inventory

This summary describes the RETC computer code for analyzing the soil water retention and hydraulic conductivity functions of unsaturated soils. These hydraulic properties are key parameters in any quantitative description of water flow into and through the unsaturated zone of soi...

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

NASA Astrophysics Data System (ADS)

Bagni, T.; Duchateau, J. L.; Breschi, M.; Devred, A.; Nijhuis, A.

2017-09-01

Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability margin of magnets. In the last decade ITER has launched a campaign for characterization of several types of NbTi and Nb3Sn CICCs comprising quench tests with a singular sine wave fast magnetic field pulse and relatively small amplitude. The stability tests, performed in the SULTAN facility, were reproduced and analyzed using two codes: JackPot-AC/DC, an electromagnetic-thermal numerical model for CICCs, developed at the University of Twente (van Lanen and Nijhuis 2010 Cryogenics 50 139-148) and multi-constant-model (MCM) (Turck and Zani 2010 Cryogenics 50 443-9), an analytical model for CICCs coupling losses. The outputs of both codes were combined with thermal, hydraulic and electric analysis of superconducting cables to predict the minimum quench energy (MQE) (Bottura et al 2000 Cryogenics 40 617-26). The experimental AC loss results were used to calibrate the JackPot and MCM models and to reproduce the energy deposited in the cable during an MQE test. The agreement between experiments and models confirm a good comprehension of the various CICCs thermal and electromagnetic phenomena. The differences between the analytical MCM and numerical JackPot approaches are discussed. The results provide a good basis for further investigation of CICC stability under plasma scenario conditions using magnetic field pulses with lower ramp rate and higher amplitude.

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

Combining Thermal And Structural Analyses

NASA Technical Reports Server (NTRS)

Winegar, Steven R.

1990-01-01

Computer code makes programs compatible so stresses and deformations calculated. Paper describes computer code combining thermal analysis with structural analysis. Called SNIP (for SINDA-NASTRAN Interfacing Program), code provides interface between finite-difference thermal model of system and finite-element structural model when no node-to-element correlation between models. Eliminates much manual work in converting temperature results of SINDA (Systems Improved Numerical Differencing Analyzer) program into thermal loads for NASTRAN (NASA Structural Analysis) program. Used to analyze concentrating reflectors for solar generation of electric power. Large thermal and structural models needed to predict distortion of surface shapes, and SNIP saves considerable time and effort in combining models.

A Computer Program for Flow-Log Analysis of Single Holes (FLASH)

USGS Publications Warehouse

Day-Lewis, F. D.; Johnson, C.D.; Paillet, Frederick L.; Halford, K.J.

2011-01-01

A new computer program, FLASH (Flow-Log Analysis of Single Holes), is presented for the analysis of borehole vertical flow logs. The code is based on an analytical solution for steady-state multilayer radial flow to a borehole. The code includes options for (1) discrete fractures and (2) multilayer aquifers. Given vertical flow profiles collected under both ambient and stressed (pumping or injection) conditions, the user can estimate fracture (or layer) transmissivities and far-field hydraulic heads. FLASH is coded in Microsoft Excel with Visual Basic for Applications routines. The code supports manual and automated model calibration. ?? 2011, The Author(s). Ground Water ?? 2011, National Ground Water Association.

Thermal Analysis of a TREAT Fuel Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Papadias, Dionissios; Wright, Arthur E.

2014-07-09

The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

Hydraulic characterization of aquifers by thermal response testing

NASA Astrophysics Data System (ADS)

Wagner, Valentin; Blum, Philipp; Bayer, Peter

2014-05-01

Temperature as a major physical quantity of the subsurface, and naturally occurring thermal anomalies are recognized as promising passive tracers to characterize the subsurface. Accelerated by the increasing popularity of geothermal energy, also active thermal field experiments have gained interest in hydrogeology. Such experiments involve artificial local ground heating or cooling. Among these, the thermal response test (TRT) is one of the most established field investigation techniques in shallow geothermal applications. It is a common method to investigate important subsurface heat transport parameters to design sustainable ground-source heat pump (GSHP) systems. During the test, the borehole heat exchanger (BHE) is heated up with a defined amount of energy by circulating a heat carrier fluid. By comparing temperature change between BHE inlet and outlet, the ability of the BHE to transfer heat or cold to the ambient ground is assessed. However, standard interpretation does not provide any insight into the governing processes of in-situ heat transfer. We utilize a groundwater advection sensitive TRT evaluation approach based on the analytical moving line source equation. It is shown that the TRT as a classical geothermal field test can also be used as a hydrogeological field test. Our approach benefits from the fact that thermal properties, such as thermal conductivity, of natural aquifers typically are much less variable than hydraulic properties, such as hydraulic conductivity. It is possible to determine a relatively small hydraulic conductivity range with our TRT evaluation approach, given realistic ranges for thermal conductivity, volumetric heat capacity, thermal dispersivity and thermal borehole resistance. The method is successfully tested on a large-scale geothermal laboratory experiment (9 m × 6 m × 4.5 m) and with a commercially performed TRT in the field scale. The laboratory experiment consists of a layered artificial aquifer, which is penetrated by a short BHE. This BHE is used to record a groundwater influenced TRT dataset. The performed field TRT is measured at a BHE located in the Upper Rhine Valley in South-West Germany, which penetrates a 68 m thick gravel aquifer with significant horizontal groundwater flow. At both sites, the derived hydraulic conductivity ranges obtained from TRT evaluation are shown to be within the ranges obtained from classical hydrogeological methods such as sieve analysis and pumping tests. This confirms that the temperature signal recorded during thermal response tests can be employed as a thermal tracer and that the evaluation of such a signal can be applied to estimate aquifer hydraulic conductivities.

Travel-time-based thermal tracer tomography

NASA Astrophysics Data System (ADS)

Somogyvári, Márk; Bayer, Peter; Brauchler, Ralf

2016-05-01

Active thermal tracer testing is a technique to get information about the flow and transport properties of an aquifer. In this paper we propose an innovative methodology using active thermal tracers in a tomographic setup to reconstruct cross-well hydraulic conductivity profiles. This is facilitated by assuming that the propagation of the injected thermal tracer is mainly controlled by advection. To reduce the effects of density and viscosity changes and thermal diffusion, early-time diagnostics are used and specific travel times of the tracer breakthrough curves are extracted. These travel times are inverted with an eikonal solver using the staggered grid method to reduce constraints from the pre-defined grid geometry and to improve the resolution. Finally, non-reliable pixels are removed from the derived hydraulic conductivity tomograms. The method is applied to successfully reconstruct cross-well profiles as well as a 3-D block of a high-resolution fluvio-aeolian aquifer analog data set. Sensitivity analysis reveals a negligible role of the injection temperature, but more attention has to be drawn to other technical parameters such as the injection rate. This is investigated in more detail through model-based testing using diverse hydraulic and thermal conditions in order to delineate the feasible range of applications for the new tomographic approach.

Advanced multiphysics coupling for LWR fuel performance analysis

DOE PAGES

Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; ...

2015-10-01

Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics,more » particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use lower length scale models such as those used in the mesoscale MARMOT code to compute average properties, e.g. swelling or thermal conductivity. These may then be used by an engineering-scale model. Examples of this type of multiscale, multiphysics modeling are shown.« less

Assessment and modification of an ion source grid design in KSTAR neutral beam system.

PubMed

Lee, Dong Won; Shin, Kyu In; Jin, Hyung Gon; Choi, Bo Guen; Kim, Tae-Seong; Jeong, Seung Ho

2014-02-01

A new 2 MW NB (Neutral Beam) ion source for supplying 3.5 MW NB heating for the KSTAR campaign was developed in 2012 and its grid was made from OFHC (Oxygen Free High Conductivity) copper with rectangular cooling channels. However, the plastic deformation such as a bulging in the plasma grid of the ion source was found during the overhaul period after the 2012 campaign. A thermal-hydraulic and a thermo-mechanical analysis using the conventional code, ANSYS, were carried out and the thermal fatigue life assessment was evaluated. It was found that the thermal fatigue life of the OFHC copper grid was about 335 cycles in case of 0.165 MW/m(2) heat flux and it gave too short fatigue life to be used as a KSTAR NB ion source grid. To overcome the limited fatigue life of the current design, the following methods were proposed in the present study: (1) changing the OHFC copper to CuCrZr, copper-alloy or (2) adopting a new design with a pure Mo metal grid and CuCrZr tubes. It is confirmed that the proposed methods meet the requirements by performing the same assessment.

ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

NASA Technical Reports Server (NTRS)

Walton, J. T.

1994-01-01

ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

75 FR 44818 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ESBWR

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-29

...-Drop and Thermal-Hydraulic Analysis for ESBWR Fuel Racks,'' NEDC- 33374P, ``Safety Analysis Report for Fuel Storage Racks Criticality Analysis for ESBWR Plants,'' NEDE-33083P, ``TRACG Application for ESBWR...

Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mays, Brian; Jackson, R. Brian

2017-03-08

The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services.more » The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.« less

Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

NASA Astrophysics Data System (ADS)

Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

2017-12-01

This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

Application of numerical methods to heat transfer and thermal stress analysis of aerospace vehicles

NASA Technical Reports Server (NTRS)

Wieting, A. R.

1979-01-01

The paper describes a thermal-structural design analysis study of a fuel-injection strut for a hydrogen-cooled scramjet engine for a supersonic transport, utilizing finite-element methodology. Applications of finite-element and finite-difference codes to the thermal-structural design-analysis of space transports and structures are discussed. The interaction between the thermal and structural analyses has led to development of finite-element thermal methodology to improve the integration between these two disciplines. The integrated thermal-structural analysis capability developed within the framework of a computer code is outlined.

An information theoretic approach to use high-fidelity codes to calibrate low-fidelity codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lewis, Allison, E-mail: lewis.allison10@gmail.com; Smith, Ralph; Williams, Brian

For many simulation models, it can be prohibitively expensive or physically infeasible to obtain a complete set of experimental data to calibrate model parameters. In such cases, one can alternatively employ validated higher-fidelity codes to generate simulated data, which can be used to calibrate the lower-fidelity code. In this paper, we employ an information-theoretic framework to determine the reduction in parameter uncertainty that is obtained by evaluating the high-fidelity code at a specific set of design conditions. These conditions are chosen sequentially, based on the amount of information that they contribute to the low-fidelity model parameters. The goal is tomore » employ Bayesian experimental design techniques to minimize the number of high-fidelity code evaluations required to accurately calibrate the low-fidelity model. We illustrate the performance of this framework using heat and diffusion examples, a 1-D kinetic neutron diffusion equation, and a particle transport model, and include initial results from the integration of the high-fidelity thermal-hydraulics code Hydra-TH with a low-fidelity exponential model for the friction correlation factor.« less

RTE: A computer code for Rocket Thermal Evaluation

NASA Technical Reports Server (NTRS)

Naraghi, Mohammad H. N.

1995-01-01

The numerical model for a rocket thermal analysis code (RTE) is discussed. RTE is a comprehensive thermal analysis code for thermal analysis of regeneratively cooled rocket engines. The input to the code consists of the composition of fuel/oxidant mixture and flow rates, chamber pressure, coolant temperature and pressure. dimensions of the engine, materials and the number of nodes in different parts of the engine. The code allows for temperature variation in axial, radial and circumferential directions. By implementing an iterative scheme, it provides nodal temperature distribution, rates of heat transfer, hot gas and coolant thermal and transport properties. The fuel/oxidant mixture ratio can be varied along the thrust chamber. This feature allows the user to incorporate a non-equilibrium model or an energy release model for the hot-gas-side. The user has the option of bypassing the hot-gas-side calculations and directly inputting the gas-side fluxes. This feature is used to link RTE to a boundary layer module for the hot-gas-side heat flux calculations.

A comparison of TSS and TRASYS in form factor calculation

NASA Technical Reports Server (NTRS)

Golliher, Eric

1993-01-01

As the workstation and personal computer become more popular than a centralized mainframe to perform thermal analysis, the methods for space vehicle thermal analysis will change. Already, many thermal analysis codes are now available for workstations, which were not in existence just five years ago. As these changes occur, some organizations will adopt the new codes and analysis techniques, while others will not. This might lead to misunderstandings between thermal shops in different organizations. If thermal analysts make an effort to understand the major differences between the new and old methods, a smoother transition to a more efficient and more versatile thermal analysis environment will be realized.

Finite element based electric motor design optimization

NASA Technical Reports Server (NTRS)

Campbell, C. Warren

1993-01-01

The purpose of this effort was to develop a finite element code for the analysis and design of permanent magnet electric motors. These motors would drive electromechanical actuators in advanced rocket engines. The actuators would control fuel valves and thrust vector control systems. Refurbishing the hydraulic systems of the Space Shuttle after each flight is costly and time consuming. Electromechanical actuators could replace hydraulics, improve system reliability, and reduce down time.

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

DOE PAGES

Brown, C. S.; Zhang, Hongbin

2016-05-24

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anh Bui; Nam Dinh; Brian Williams

In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Suchmore » sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable quantitative assessment of the CASL modeling of Crud-Induced Power Shift (CIPS) phenomenon, in particular, and the CASL advanced predictive capabilities, in general. This report is prepared for the Department of Energy’s Consortium for Advanced Simulation of LWRs program’s VUQ Focus Area.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

DOE PAGES

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

2018-03-20

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Estimating spatially distributed soil texture using time series of thermal remote sensing - a case study in central Europe

NASA Astrophysics Data System (ADS)

Müller, Benjamin; Bernhardt, Matthias; Jackisch, Conrad; Schulz, Karsten

2016-09-01

For understanding water and solute transport processes, knowledge about the respective hydraulic properties is necessary. Commonly, hydraulic parameters are estimated via pedo-transfer functions using soil texture data to avoid cost-intensive measurements of hydraulic parameters in the laboratory. Therefore, current soil texture information is only available at a coarse spatial resolution of 250 to 1000 m. Here, a method is presented to derive high-resolution (15 m) spatial topsoil texture patterns for the meso-scale Attert catchment (Luxembourg, 288 km2) from 28 images of ASTER (advanced spaceborne thermal emission and reflection radiometer) thermal remote sensing. A principle component analysis of the images reveals the most dominant thermal patterns (principle components, PCs) that are related to 212 fractional soil texture samples. Within a multiple linear regression framework, distributed soil texture information is estimated and related uncertainties are assessed. An overall root mean squared error (RMSE) of 12.7 percentage points (pp) lies well within and even below the range of recent studies on soil texture estimation, while requiring sparser sample setups and a less diverse set of basic spatial input. This approach will improve the generation of spatially distributed topsoil maps, particularly for hydrologic modeling purposes, and will expand the usage of thermal remote sensing products.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, C. S.; Zhang, Hongbin

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Statistical core design methodology using the VIPRE thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lloyd, M.W.; Feltus, M.A.

1994-12-31

This Penn State Statistical Core Design Methodology (PSSCDM) is unique because it not only includes the EPRI correlation/test data standard deviation but also the computational uncertainty for the VIPRE code model and the new composite box design correlation. The resultant PSSCDM equation mimics the EPRI DNBR correlation results well, with an uncertainty of 0.0389. The combined uncertainty yields a new DNBR limit of 1.18 that will provide more plant operational flexibility. This methodology and its associated correlation and uniqe coefficients are for a very particular VIPRE model; thus, the correlation will be specifically linked with the lumped channel and subchannelmore » layout. The results of this research and methodology, however, can be applied to plant-specific VIPRE models.« less

Temperature distribution by the effect of groundwater flow in an aquifer thermal energy storage system model

NASA Astrophysics Data System (ADS)

Shim, B.

2005-12-01

Aquifer thermal energy storage (ATES) can be a cost-effective and renewable energy source, depending on site-specific thermohydraulic conditions. To design an effective ATES system, the understanding of thermohydraulic processes is necessary. The heat transfer phenomena of an aquifer heat storage system are simulated with the scenario of heat pump operation of pumping and waste water reinjection in a two layered confined aquifer model having the effect of groundwater movement. Temperature distribution of the aquifer model is generated, and hydraulic heads and temperature variations are monitored at both wells during simulation days. The average groundwater velocities are determined with two assumed hydraulic gradients set by boundary conditions, and the effect of groundwater flow are shown at the generated thermal distributions at three different depth slices. The generated temperature contour lines at the hydraulic gradient of 0.001 are shaped circular, and the center is moved less than 5 m to the east in 365 days. However at the hydraulic gradient of 0.01, the contour centers of the east well at each depth slice are moved near the east boundary and the movement of temperature distribution is increased at the lower aquifer. By the analysis of thermal interference data between two wells the efficiency of a heat pump operation model is validated, and the variation of heads is monitored at injection, pumping and stabilized state. The thermal efficiency of the ATES system model is represented as highly depended on groundwater flow velocity and direction. Therefore the hydrogeologic condition for the system site should be carefully surveyed.

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

Analysis of the influence of the heat transfer phenomena on the late phase of the ThAI Iod-12 test

NASA Astrophysics Data System (ADS)

Gonfiotti, B.; Paci, S.

2014-11-01

Iodine is one of the major contributors to the source term during a severe accident in a Nuclear Power Plant for its volatility and high radiological consequences. Therefore, large efforts have been made to describe the Iodine behaviour during an accident, especially in the containment system. Due to the lack of experimental data, in the last years many attempts were carried out to fill the gaps on the knowledge of Iodine behaviour. In this framework, two tests (ThAI Iod-11 and Iod-12) were carried out inside a multi-compartment steel vessel. A quite complex transient characterizes these two tests; therefore they are also suitable for thermal- hydraulic benchmarks. The two tests were originally released for a benchmark exercise during the SARNET2 EU Project. At the end of this benchmark a report covering the main findings was issued, stating that the common codes employed in SA studies were able to simulate the tests but with large discrepancies. The present work is then related to the application of the new versions of ASTEC and MELCOR codes with the aim of carry out a new code-to-code comparison vs. ThAI Iod-12 experimental data, focusing on the influence of the heat exchanges with the outer environment, which seems to be one of the most challenging issues to cope with.

Hydro-mechanical coupled simulation of hydraulic fracturing using the eXtended Finite Element Method (XFEM)

NASA Astrophysics Data System (ADS)

Youn, Dong Joon

This thesis presents the development and validation of an advanced hydro-mechanical coupled finite element program analyzing hydraulic fracture propagation within unconventional hydrocarbon formations under various conditions. The realistic modeling of hydraulic fracturing is necessarily required to improve the understanding and efficiency of the stimulation technique. Such modeling remains highly challenging, however, due to factors including the complexity of fracture propagation mechanisms, the coupled behavior of fracture displacement and fluid pressure, the interactions between pre-existing natural and initiated hydraulic fractures and the formation heterogeneity of the target reservoir. In this research, an eXtended Finite Element Method (XFEM) scheme is developed allowing for representation of single or multiple fracture propagations without any need for re-meshing. Also, the coupled flows through the fracture are considered in the program to account for their influence on stresses and deformations along the hydraulic fracture. In this research, a sequential coupling scheme is applied to estimate fracture aperture and fluid pressure with the XFEM. Later, the coupled XFEM program is used to estimate wellbore bottomhole pressure during fracture propagation, and the pressure variations are analyzed to determine the geometry and performance of the hydraulic fracturing as pressure leak-off test. Finally, material heterogeneity is included into the XFEM program to check the effect of random formation property distributions to the hydraulic fracture geometry. Random field theory is used to create the random realization of the material heterogeneity with the consideration of mean, standard deviation, and property correlation length. These analyses lead to probabilistic information on the response of unconventional reservoirs and offer a more scientific approach regarding risk management for the unconventional reservoir stimulation. The new stochastic approach combining XFEM and random field is named as eXtended Random Finite Element Method (XRFEM). All the numerical analysis codes in this thesis are written in Fortran 2003, and these codes are applicable as a series of sub-modules within a suite of finite element codes developed by Smith and Griffiths (2004).

Computer codes for thermal analysis of a solid rocket motor nozzle

NASA Technical Reports Server (NTRS)

Chauhan, Rajinder Singh

1988-01-01

A number of computer codes are available for performing thermal analysis of solid rocket motor nozzles. Aerotherm Chemical Equilibrium (ACE) computer program can be used to perform one-dimensional gas expansion to determine the state of the gas at each location of a nozzle. The ACE outputs can be used as input to a computer program called Momentum/Energy Integral Technique (MEIT) for predicting boundary layer development development, shear, and heating on the surface of the nozzle. The output from MEIT can be used as input to another computer program called Aerotherm Charring Material Thermal Response and Ablation Program (CMA). This program is used to calculate oblation or decomposition response of the nozzle material. A code called Failure Analysis Nonlinear Thermal and Structural Integrated Code (FANTASTIC) is also likely to be used for performing thermal analysis of solid rocket motor nozzles after the program is duly verified. A part of the verification work on FANTASTIC was done by using one and two dimension heat transfer examples with known answers. An attempt was made to prepare input for performing thermal analysis of the CCT nozzle using the FANTASTIC computer code. The CCT nozzle problem will first be solved by using ACE, MEIT, and CMA. The same problem will then be solved using FANTASTIC. These results will then be compared for verification of FANTASTIC.

Development and Assessment of CTF for Pin-resolved BWR Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

MOOSE Implementation of MAMBA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Galloway, Jack; Matthews, Topher

The development of MAMBA is targeted at capturing both core wide CRUD induced power shifts (CIPS) as well as pin-­level CRUD induced localized corrosion (CILC). Both CIPS and CILC require some sort of information from thermal-­hydraulic, neutronics, and fuel performance codes, although the degree of coupling is different for the two effects. Since CIPS necessarily requires a core-­wide power distribution solve, it requires tight coupling with a neutronics code. Conversely, CIPS tends to be an individual pin phenomenon, requiring tight coupling a fuel performance code. As efforts are now focused on coupling MAMBA within the VERA suite, a natural separationmore » has surfaced in which a FORTRAN rewrite of MAMBA is optimal for VERA integration to capture CIPS behavior, while a CILC focused calculation would benefit from a tight coupling with BISON, motivating a MOOSE version of MAMBA.« less

Posttest RELAP5 simulations of the Semiscale S-UT series experiments. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leonard, M.T.

The RELAP5/MOD1 computer code was used to perform posttest calculations, simulating six experiments, run in the Semiscale Mod-2A facility, investigating the effects of upper head injection on small break transient behavior. The results of these calculations and corresponding test data are presented in this report. An evaluation is made of the capability of RELAP5 to calculate the thermal-hydraulic response of the Mod-2A system over a spectrum of break sizes, with and without the use of upper head injection.

TEMPEST. Transient 3-D Thermal-Hydraulic

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eyler, L.L.

TEMPEST is a transient, three-dimensional, hydrothermal program that is designed to analyze a range of coupled fluid dynamic and heat transfer systems of particular interest to the Fast Breeder Reactor (FBR) thermal-hydraulic design community. The full three-dimensional, time-dependent equations of motion, continuity, and heat transport are solved for either laminar or turbulent fluid flow, including heat diffusion and generation in both solid and liquid materials. The equations governing mass, momentum, and energy conservation for incompressible flows and small density variations (Boussinesq approximation) are solved using finite-difference techniques. Analyses may be conducted in either cylindrical or Cartesian coordinate systems. Turbulence ismore » treated using a two-equation model. Two auxiliary plotting programs, SEQUEL and MANPLOT, for use with TEMPEST output are included. SEQUEL may be operated in batch or interactive mode; it generates data required for vector plots, contour plots of scalar quantities, line plots, grid and boundary plots, and time-history plots. MANPLOT reads the SEQUEL-generated data and creates the hardcopy plots. TEMPEST can be a valuable hydrothermal design analysis tool in areas outside the intended FBR thermal-hydraulic design community.« less

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Modeling the effects of hydraulic stimulation on geothermal reservoirs

NASA Astrophysics Data System (ADS)

De Simone, Silvia; Vilarrasa, Victor; Carrera, Jesús; Alcolea, Andrés; Meier, Peter

2013-04-01

Geothermal energy represents a huge power source that can provide clean energy in potentially unlimited supply. When designing geothermal energy production from deep hot rocks, permeability is considered to control the economic efficiency of the heat extraction operations. In fact, a high permeability heat exchanger is required to achieve a cost-competitive power generation. The typical procedure entails intercepting naturally fractured rocks and enhancing their permeability by means of stimulation. Hydraulic stimulation is the most widely used method. It involves the massive injection of a large volume of water at high flow rates to increase the downhole pore pressure. This overpressure reduces the effective stresses, which tends to induce shearing along the fracture planes. In this way permeability is enhanced due to dilatancy, especially in the direction perpendicular to shear. These processes usually trigger microseismic events, which are sometimes of sufficient magnitude to be felt by the local population. This causes a negative impact on the local population and may compromise the continuation of the project. Hence, understanding the mechanisms triggering these induced micro-earthquakes is important to properly design and manage geothermal stimulation and operations so as to prevent them. We analyzed the thermo-hydro-mechanical response of a fractured deep rock mass subjected to hydraulic stimulation. Considering that seismicity is triggered when failure condition are reached, we studied the variation of the stress regime due to the hydraulic and thermal perturbations during fluid injection. Starting with a simplified model with constant permeability fault zones, more sophisticated schemes are considered to simulate the behavior of the discontinuity zones, including permeability variation associated to temperature, pressure and stress regime changes. Numerical simulations are performed using the finite element numerical code CODE_BRIGHT, which allows to solve fully coupled thermo-hydro-mechanical problems. Results allowed to estimate the impact of the hydraulic stimulation on the overall behavior.

Recovery Act. Development and Validation of an Advanced Stimulation Prediction Model for Enhanced Geothermal System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gutierrez, Marte

The research project aims to develop and validate an advanced computer model that can be used in the planning and design of stimulation techniques to create engineered reservoirs for Enhanced Geothermal Systems. The specific objectives of the proposal are to: 1) Develop a true three-dimensional hydro-thermal fracturing simulator that is particularly suited for EGS reservoir creation. 2) Perform laboratory scale model tests of hydraulic fracturing and proppant flow/transport using a polyaxial loading device, and use the laboratory results to test and validate the 3D simulator. 3) Perform discrete element/particulate modeling of proppant transport in hydraulic fractures, and use the resultsmore » to improve understand of proppant flow and transport. 4) Test and validate the 3D hydro-thermal fracturing simulator against case histories of EGS energy production. 5) Develop a plan to commercialize the 3D fracturing and proppant flow/transport simulator. The project is expected to yield several specific results and benefits. Major technical products from the proposal include: 1) A true-3D hydro-thermal fracturing computer code that is particularly suited to EGS, 2) Documented results of scale model tests on hydro-thermal fracturing and fracture propping in an analogue crystalline rock, 3) Documented procedures and results of discrete element/particulate modeling of flow and transport of proppants for EGS applications, and 4) Database of monitoring data, with focus of Acoustic Emissions (AE) from lab scale modeling and field case histories of EGS reservoir creation.« less

Recovery Act. Development and Validation of an Advanced Stimulation Prediction Model for Enhanced Geothermal Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gutierrez, Marte

2013-12-31

This research project aims to develop and validate an advanced computer model that can be used in the planning and design of stimulation techniques to create engineered reservoirs for Enhanced Geothermal Systems. The specific objectives of the proposal are to; Develop a true three-dimensional hydro-thermal fracturing simulator that is particularly suited for EGS reservoir creation; Perform laboratory scale model tests of hydraulic fracturing and proppant flow/transport using a polyaxial loading device, and use the laboratory results to test and validate the 3D simulator; Perform discrete element/particulate modeling of proppant transport in hydraulic fractures, and use the results to improve understandmore » of proppant flow and transport; Test and validate the 3D hydro-thermal fracturing simulator against case histories of EGS energy production; and Develop a plan to commercialize the 3D fracturing and proppant flow/transport simulator. The project is expected to yield several specific results and benefits. Major technical products from the proposal include; A true-3D hydro-thermal fracturing computer code that is particularly suited to EGS; Documented results of scale model tests on hydro-thermal fracturing and fracture propping in an analogue crystalline rock; Documented procedures and results of discrete element/particulate modeling of flow and transport of proppants for EGS applications; and Database of monitoring data, with focus of Acoustic Emissions (AE) from lab scale modeling and field case histories of EGS reservoir creation.« less

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo

The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the descriptionmore » of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.« less

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE PAGES

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin; ...

2017-12-22

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD MHTGR-350 Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom

2014-04-01

The INL PHISICS code system consists of three modules providing improved core simulation capability: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been finalized, and as part of the code verification and validation program the exercises defined for Phase I of the OECD/NEA MHTGR 350 MW Benchmark were completed. This paper provides an overview of the MHTGR Benchmark, and presents selected results of the three steady state exercises 1-3 defined for Phase I. For Exercise 1,more » a stand-alone steady-state neutronics solution for an End of Equilibrium Cycle Modular High Temperature Reactor (MHTGR) was calculated with INSTANT, using the provided geometry, material descriptions, and detailed cross-section libraries. Exercise 2 required the modeling of a stand-alone thermal fluids solution. The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 combined the first two exercises in a coupled neutronics and thermal fluids solution, and the coupled code suite PHISICS/RELAP5-3D was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of the traditional RELAP5-3D “ring” model approach vs. a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity of the block model is illustrated with comparison results on the temperature, power density and flux distributions, and the typical under-predictions produced by the ring model approach are highlighted.« less

Aerothermo-Structural Analysis of Low Cost Composite Nozzle/Inlet Components

NASA Technical Reports Server (NTRS)

Shivakumar, Kuwigai; Challa, Preeli; Sree, Dave; Reddy, D.

1999-01-01

This research is a cooperative effort among the Turbomachinery and Propulsion Division of NASA Glenn, CCMR of NC A&T State University, and the Tuskegee University. The NC A&T is the lead center and Tuskegee University is the participating institution. Objectives of the research were to develop an integrated aerodynamic, thermal and structural analysis code for design of aircraft engine components, such as, nozzles and inlets made of textile composites; conduct design studies on typical inlets for hypersonic transportation vehicles and setup standards test examples and finally manufacture a scaled down composite inlet. These objectives are accomplished through the following seven tasks: (1) identify the relevant public domain codes for all three types of analysis; (2) evaluate the codes for the accuracy of results and computational efficiency; (3) develop aero-thermal and thermal structural mapping algorithms; (4) integrate all the codes into one single code; (5) write a graphical user interface to improve the user friendliness of the code; (6) conduct test studies for rocket based combined-cycle engine inlet; and finally (7) fabricate a demonstration inlet model using textile preform composites. Tasks one, two and six are being pursued. Selected and evaluated NPARC for flow field analysis, CSTEM for in-depth thermal analysis of inlets and nozzles and FRAC3D for stress analysis. These codes have been independently verified for accuracy and performance. In addition, graphical user interface based on micromechanics analysis for laminated as well as textile composites was developed. Demonstration of this code will be made at the conference. A rocket based combined cycle engine was selected for test studies. Flow field analysis of various inlet geometries were studied. Integration of codes is being continued. The codes developed are being applied to a candidate example of trailblazer engine proposed for space transportation. A successful development of the code will provide a simpler, faster and user-friendly tool for conducting design studies of aircraft and spacecraft engines, applicable in high speed civil transport and space missions.

Thermal Stratification Analysis for Sodium Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

Burner liner thermal/structural load modeling: TRANCITS program user's manual

NASA Technical Reports Server (NTRS)

Maffeo, R.

1985-01-01

Transfer Analysis Code to Interface Thermal/Structural Problems (TRANCITS) is discussed. The TRANCITS code satisfies all the objectives for transferring thermal data between heat transfer and structural models of combustor liners and it can be used as a generic thermal translator between heat transfer and stress models of any component, regardless of the geometry. The TRANCITS can accurately and efficiently convert the temperature distributions predicted by the heat transfer programs to those required by the stress codes. It can be used for both linear and nonlinear structural codes and can produce nodal temperatures, elemental centroid temperatures, or elemental Gauss point temperatures. The thermal output of both the MARC and SINDA heat transfer codes can be interfaced directly with TRANCITS, and it will automatically produce stress model codes formatted for NASTRAN and MARC. Any thermal program and structural program can be interfaced by using the neutral input and output forms supported by TRANCITS.

SHARP User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yu, Y. Q.; Shemon, E. R.; Thomas, J. W.

SHARP is an advanced modeling and simulation toolkit for the analysis of nuclear reactors. It is comprised of several components including physical modeling tools, tools to integrate the physics codes for multi-physics analyses, and a set of tools to couple the codes within the MOAB framework. Physics modules currently include the neutronics code PROTEUS, the thermal-hydraulics code Nek5000, and the structural mechanics code Diablo. This manual focuses on performing multi-physics calculations with the SHARP ToolKit. Manuals for the three individual physics modules are available with the SHARP distribution to help the user to either carry out the primary multi-physics calculationmore » with basic knowledge or perform further advanced development with in-depth knowledge of these codes. This manual provides step-by-step instructions on employing SHARP, including how to download and install the code, how to build the drivers for a test case, how to perform a calculation and how to visualize the results. Since SHARP has some specific library and environment dependencies, it is highly recommended that the user read this manual prior to installing SHARP. Verification tests cases are included to check proper installation of each module. It is suggested that the new user should first follow the step-by-step instructions provided for a test problem in this manual to understand the basic procedure of using SHARP before using SHARP for his/her own analysis. Both reference output and scripts are provided along with the test cases in order to verify correct installation and execution of the SHARP package. At the end of this manual, detailed instructions are provided on how to create a new test case so that user can perform novel multi-physics calculations with SHARP. Frequently asked questions are listed at the end of this manual to help the user to troubleshoot issues.« less

Research subjects for analytical estimation of core degradation at Fukushima-Daiichi nuclear power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nagase, F.; Ishikawa, J.; Kurata, M.

2013-07-01

Estimation of the accident progress and status inside the pressure vessels (RPV) and primary containment vessels (PCV) is required for appropriate conductance of decommissioning in the Fukushima-Daiichi NPP. For that, it is necessary to obtain additional experimental data and revised models for the estimation using computer codes with increased accuracies. The Japan Atomic Energy Agency (JAEA) has selected phenomena to be reviewed and developed, considering previously obtained information, conditions specific to the Fukushima-Daiichi NPP accident, and recent progress of experimental and analytical technologies. As a result, research and development items have been picked up in terms of thermal-hydraulic behavior inmore » the RPV and PCV, progression of fuel bundle degradation, failure of the lower head of RPV, and analysis of the accident. This paper introduces the selected phenomena to be reviewed and developed, research plans and recent results from the JAEA's corresponding research programs. (authors)« less

Applications of the RELAP5 code to the station blackout transients at the Browns Ferry Unit One Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schultz, R.R.; Wagoner, S.R.

1983-01-01

As a part of the charter of the Severe Accident Sequence Analysis (SASA) Program, station blackout transients have been analyzed using a RELAP5 model of the Browns Ferry Unit 1 Plant. The task was conducted as a partial fulfillment of the needs of the US Nuclear Regulatory Commission in examining the Unresolved Safety Issue A-44: Station Blackout (1) the station blackout transients were examined (a) to define the equipment needed to maintain a well cooled core, (b) to determine when core uncovery would occur given equipment failure, and (c) to characterize the behavior of the vessel thermal-hydraulics during the stationmore » blackout transients (in part as the plant operator would see it). These items are discussed in the paper. Conclusions and observations specific to the station blackout are presented.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS Advanced Reactor is a 640-MW(e) pressurized-water reactor developed by Asea Brown Boveri. A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity normally is controlled by the boron concentration in the coolant and the temperature of the moderator coolant. Analyses of five initiating events have been completed on the basis of calculations performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. The initiating events analyzed are (1) reactor scram, (2) loss of off-site power (3) main steam-line break, (4) small-break loss of coolant, and (5) large-break loss of coolant. Inmore » addition to the baseline calculation for each sequence, sensitivity studies were performed to explore the response of the PIUS reactor to severe off-normal conditions having a very low probability of occurrence. The sensitivity studies provide insights into the robustness of the design.« less

77 FR 5063 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-01

... Subcommittee on Thermal-Hydraulics Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal-Hydraulics... Regulatory Guide (1.79), ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water... Water Reactors.'' The Subcommittee will hear presentations by and hold discussions with the NRC staff...

Characterization of an alluvial aquifer with thermal tracer tomography

NASA Astrophysics Data System (ADS)

Somogyvári, Márk; Bayer, Peter

2017-04-01

In the summer of 2015, a series of thermal tracer tests was performed at the Widen field site in northeast Switzerland. At this site numerous hydraulic, tracer, geophysical and hydrogeophysical field tests have been conducted in the past to investigate a shallow alluvial aquifer. The goals of the campaign in 2015 were to design a cost-effective thermal tracer tomography setup and to validate the concept of travel time-based thermal tracer tomography under field conditions. Thermal tracer tomography uses repeated thermal tracer injections with different injection depths and distributed temperature measurements to map the hydraulic conductivity distribution of a heterogeneous aquifer. The tracer application was designed with minimal experimental time and cost. Water was heated in inflatable swimming pools using direct sunlight of the warm summer days, and it was injected as low temperature pulses in a well. Because of the small amount of injected heat, no long recovery times were required between the repeated heat tracer injections and every test started from natural thermal conditions. At Widen, four thermal tracer tests were performed during a period of three days. Temperatures were measured in one downgradient well using a distributed temperature measurement system installed at seven depth points. Totally 12 temperature breakthrough curves were collected. Travel time based tomographic inversion assumes that thermal transport is dominated by advection and the travel time of the thermal tracer can be related to the hydraulic conductivities of the aquifer. This assumption is valid in many shallow porous aquifers where the groundwater flow is fast. In our application, the travel time problem was treated by a tomographic solver, analogous to seismic tomography, to derive the hydraulic conductivity distribution. At the test site, a two-dimensional cross-well hydraulic conductivity profile was reconstructed with the travel time based inversion. The reconstructed profile corresponds well with the findings of the earlier hydraulic and geophysical experiments at the site.

Discrimination of soil hydraulic properties by combined thermal infrared and microwave remote sensing

NASA Technical Reports Server (NTRS)

Vandegriend, A. A.; Oneill, P. E.

1986-01-01

Using the De Vries models for thermal conductivity and heat capacity, thermal inertia was determined as a function of soil moisture for 12 classes of soil types ranging from sand to clay. A coupled heat and moisture balance model was used to describe the thermal behavior of the top soil, while microwave remote sensing was used to estimate the soil moisture content of the same top soil. Soil hydraulic parameters are found to be very highly correlated with the combination of soil moisture content and thermal inertia at the same moisture content. Therefore, a remotely sensed estimate of the thermal behavior of the soil from diurnal soil temperature observations and an independent remotely sensed estimate of soil moisture content gives the possibility of estimating soil hydraulic properties by remote sensing.

A THC Simulator for Modeling Fluid-Rock Interactions

NASA Astrophysics Data System (ADS)

Hamidi, Sahar; Galvan, Boris; Heinze, Thomas; Miller, Stephen

2014-05-01

Fluid-rock interactions play an essential role in many earth processes, from a likely influence on earthquake nucleation and aftershocks, to enhanced geothermal system, carbon capture and storage (CCS), and underground nuclear waste repositories. In THC models, two-way interactions between different processes (thermal, hydraulic and chemical) are present. Fluid flow influences the permeability of the rock especially if chemical reactions are taken into account. On one hand solute concentration influences fluid properties while, on the other hand, heat can affect further chemical reactions. Estimating heat production from a naturally fractured geothermal systems remains a complex problem. Previous works are typically based on a local thermal equilibrium assumption and rarely consider the salinity. The dissolved salt in fluid affects the hydro- and thermodynamical behavior of the system by changing the hydraulic properties of the circulating fluid. Coupled thermal-hydraulic-chemical models (THC) are important for investigating these processes, but what is needed is a coupling to mechanics to result in THMC models. Although similar models currently exist (e.g. PFLOTRAN), our objective here is to develop algorithms for implementation using the Graphics Processing Unit (GPU) computer architecture to be run on GPU clusters. To that aim, we present a two-dimensional numerical simulation of a fully coupled non-isothermal non-reactive solute flow. The thermal part of the simulation models heat transfer processes for either local thermal equilibrium or nonequilibrium cases, and coupled to a non-reactive mass transfer described by a non-linear diffusion/dispersion model. The flow process of the model includes a non-linear Darcian flow for either saturated or unsaturated scenarios. For the unsaturated case, we use the Richards' approximation for a mixture of liquid and gas phases. Relative permeability and capillary pressure are determined by the van Genuchten relations. Permeability of rock is controlled by porosity, which is itself related to effective stress. The theoretical model is solved using explicit finite differences, and runs in parallel mode with OpenMP. The code is fully modular so that any combination of current THC processes, one- and two-phase, can be chosen. Future developments will include dissolution and precipitation of chemical components in addition to chemical erosion.

75 FR 51499 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-20

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulics Phenomena The ACRS Subcommittee on Thermal Hydraulics Phenomena will hold a meeting on September 7, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting...

75 FR 57536 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulic Phenomena The ACRS Subcommittee on Thermal Hydraulic Phenomena will hold a meeting on October 18, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting...

76 FR 53979 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-30

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulics Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal Hydraulics... Revision 4 to Regulatory Guide 1.82, ``Water Sources for Long-Term Recirculation Cooling Following a Loss...

Multi-Region Boundary Element Analysis for Coupled Thermal-Fracturing Processes in Geomaterials

NASA Astrophysics Data System (ADS)

Shen, Baotang; Kim, Hyung-Mok; Park, Eui-Seob; Kim, Taek-Kon; Wuttke, Manfred W.; Rinne, Mikael; Backers, Tobias; Stephansson, Ove

2013-01-01

This paper describes a boundary element code development on coupled thermal-mechanical processes of rock fracture propagation. The code development was based on the fracture mechanics code FRACOD that has previously been developed by Shen and Stephansson (Int J Eng Fracture Mech 47:177-189, 1993) and FRACOM (A fracture propagation code—FRACOD, User's manual. FRACOM Ltd. 2002) and simulates complex fracture propagation in rocks governed by both tensile and shear mechanisms. For the coupled thermal-fracturing analysis, an indirect boundary element method, namely the fictitious heat source method, was implemented in FRACOD to simulate the temperature change and thermal stresses in rocks. This indirect method is particularly suitable for the thermal-fracturing coupling in FRACOD where the displacement discontinuity method is used for mechanical simulation. The coupled code was also extended to simulate multiple region problems in which rock mass, concrete linings and insulation layers with different thermal and mechanical properties were present. Both verification and application cases were presented where a point heat source in a 2D infinite medium and a pilot LNG underground cavern were solved and studied using the coupled code. Good agreement was observed between the simulation results, analytical solutions and in situ measurements which validates an applicability of the developed coupled code.

Cyclic injection, storage, and withdrawal of heated water in a sandstone aquifer at St. Paul, Minnesota: Field observations, preliminary model analysis, and aquifer thermal efficiency

USGS Publications Warehouse

Miller, Robert T.

1989-01-01

The Franconia-Ironton-Galesville aquifer is a consolidated sandstone, approximately 60 m thick, the top of which is approximately 180 m below the land surface. It is confined above by the St. Lawrence Formation--a dolomitic sandstone 8-m thick--and below by the Eau Claire Formation--a shale 30-m thick. Initial hydraulic testing with inflatable packers indicated that the aquifer has four hydraulic zones with distinctly different values of relative horizontal hydraulic conductivity. The thickness of each zone was determined by correlating data from geophysical logs, core samples, and the inflatablepacker tests.

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Modeling the effects of the variability of temperature-related dynamic viscosity on the thermal-affected zone of groundwater heat-pump systems

NASA Astrophysics Data System (ADS)

Lo Russo, Stefano; Taddia, Glenda; Cerino Abdin, Elena

2018-06-01

Thermal perturbation in the subsurface produced in an open-loop groundwater heat pump (GWHP) plant is a complex transport phenomenon affected by several factors, including the exploited aquifer's hydrogeological and thermal characteristics, well construction features, and the temporal dynamics of the plant's groundwater abstraction and reinjection system. Hydraulic conductivity has a major influence on heat transport because plume propagation, which occurs primarily through advection, tends to degrade following conductive heat transport and convection within moving water. Hydraulic conductivity is, in turn, influenced by water reinjection because the dynamic viscosity of groundwater varies with temperature. This paper reports on a computational analysis conducted using FEFLOW software to quantify how the thermal-affected zone (TAZ) is influenced by the variation in dynamic viscosity due to reinjected groundwater in a well-doublet scheme. The modeling results demonstrate non-negligible groundwater dynamic-viscosity variation that affects thermal plume propagation in the aquifer. This influence on TAZ calculation was enhanced for aquifers with high intrinsic permeability and/or substantial temperature differences between abstracted and post-heat-pump-reinjected groundwater.

Modeling the effects of the variability of temperature-related dynamic viscosity on the thermal-affected zone of groundwater heat-pump systems

NASA Astrophysics Data System (ADS)

Lo Russo, Stefano; Taddia, Glenda; Cerino Abdin, Elena

2018-01-01

Thermal perturbation in the subsurface produced in an open-loop groundwater heat pump (GWHP) plant is a complex transport phenomenon affected by several factors, including the exploited aquifer's hydrogeological and thermal characteristics, well construction features, and the temporal dynamics of the plant's groundwater abstraction and reinjection system. Hydraulic conductivity has a major influence on heat transport because plume propagation, which occurs primarily through advection, tends to degrade following conductive heat transport and convection within moving water. Hydraulic conductivity is, in turn, influenced by water reinjection because the dynamic viscosity of groundwater varies with temperature. This paper reports on a computational analysis conducted using FEFLOW software to quantify how the thermal-affected zone (TAZ) is influenced by the variation in dynamic viscosity due to reinjected groundwater in a well-doublet scheme. The modeling results demonstrate non-negligible groundwater dynamic-viscosity variation that affects thermal plume propagation in the aquifer. This influence on TAZ calculation was enhanced for aquifers with high intrinsic permeability and/or substantial temperature differences between abstracted and post-heat-pump-reinjected groundwater.

RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom

2012-06-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requiresmore » participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.« less

RELAP5-3D results for phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, G.; Epiney, A. S.

2012-07-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requiresmore » participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2. (authors)« less

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1977-06-01

This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

Conceptual design and cost analysis of hydraulic output unit for 15 kW free-piston Stirling engine

NASA Technical Reports Server (NTRS)

White, M. A.

1982-01-01

A long-life hydraulic converter with unique features was conceptually designed to interface with a specified 15 kW(e) free-piston Stirling engine in a solar thermal dish application. Hydraulic fluid at 34.5 MPa (5000 psi) is produced to drive a conventional hydraulic motor and rotary alternator. Efficiency of the low-maintenance converter design was calculated at 93.5% for a counterbalanced version and 97.0% without the counterbalance feature. If the converter were coupled to a Stirling engine with design parameters more typcial of high-technology Stirling engines, counterbalanced converter efficiency could be increased to 99.6%. Dynamic computer simulation studies were conducted to evaluate performance and system sensitivities. Production costs of the complete Stirling hydraulic/electric power system were evaluated at $6506 which compared with $8746 for an alternative Stirling engine/linear alternator system.

77 FR 24745 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-25

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Thermal Hydraulic Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal Hydraulic Phenomena will hold a meeting on May 8-9, 2012, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The...

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

An anisotropic numerical model for thermal hydraulic analyses: application to liquid metal flow in fuel assemblies

NASA Astrophysics Data System (ADS)

Vitillo, F.; Vitale Di Maio, D.; Galati, C.; Caruso, G.

2015-11-01

A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

Thermal and orbital analysis of Earth monitoring Sun-synchronous space experiments

NASA Technical Reports Server (NTRS)

Killough, Brian D.

1990-01-01

The fundamentals of an Earth monitoring Sun-synchronous orbit are presented. A Sun-synchronous Orbit Analysis Program (SOAP) was developed to calculate orbital parameters for an entire year. The output from this program provides the required input data for the TRASYS thermal radiation computer code, which in turn computes the infrared, solar and Earth albedo heat fluxes incident on a space experiment. Direct incident heat fluxes can be used as input to a generalized thermal analyzer program to size radiators and predict instrument operating temperatures. The SOAP computer code and its application to the thermal analysis methodology presented, should prove useful to the thermal engineer during the design phases of Earth monitoring Sun-synchronous space experiments.

Design and Analysis of an Axial Bypass Compressor Blade in a Supercritical CO2 Gas Turbine

NASA Astrophysics Data System (ADS)

Ishizuka, Takao; Muto, Yasushi; Aritomi, Masanori; Tsuzuki, Nobuyoshi; Kikura, Hiroshige

A supercritical carbon dioxide gas turbine can generate power at a high cycle thermal efficiency, even at modest temperatures of 500-550°C. Consequently, a more reliable and economically advantageous power generation system is achieved by coupling with a Na-cooled fast reactor. This paper mainly describes the bypass compressor (a key component) design and thermal hydraulic analysis using CFD (with FLUENT code). Fluid conditions of the bypass compressor are determined by the cycle calculation of this system. Aerodynamic design was conducted using the loss model described by Cohen et al., which enables the use of several stages while providing total adiabatic efficiency of 21 and 87%, respectivly. Blade shapes were prepared based on flow angles and chord length obtained for the aerodynamic design. In the CFD analysis, the calculated value of the mass flow rate for each stage was adjusted to that of the design. The value of the design outlet pressure was reached at stage No. 16, which is fewer stages than that for design, No. 21. The difference between these stage numbers is attributed to the three-dimensional effect in design. If these effects are eliminated, then the design calculation yields an almost identical number of stages. Therefore, it was concluded that the existing design method is applicable to the supercritical CO2 bypass compressor. Furthermore, CFD analysis appears to be an effective aerodynamic design tool, but these conclusions should be verified experimentally.

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, J. S.; Cheng, L. Y.; Diamond, D.

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enrichedmore » uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.« less

Development of Switchable Polarity Solvent Draw Solutes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, Aaron D.

Results of a computational fluid dynamic (CFD) study of flow and heat transfer in a printed circuit heat exchanger (PCHE) geometry are presented. CFD results obtained from a two-plate model are compared to corresponding experimental results for the validation. This process provides the basis for further application of the CFD code to PCHE design and performance analysis in a variety of internal flow geometries. As a part of the code verification and validation (V&V) process, CFD simulation of a single semicircular straight channel under laminar isothermal conditions was also performed and compared to theoretical results. This comparison yielded excellent agreementmore » with the theoretical values. The two-plate CFD model based on the experimental PCHE design overestimated the effectiveness and underestimated the pressure drop. However, it is found that the discrepancy between the CFD result and experimental data was mainly caused by the uncertainty in the geometry of heat exchanger during the fabrication. The CFD results obtained using a slightly smaller channel diameter yielded good agreement with the experimental data. A separate investigation revealed that the average channel diameter of the OSU PCHE after the diffusion-bonding was 1.93 mm on the cold fluid side and 1.90 mm on the hot fluid side which are both smaller than the nominal design value. Consequently, the CFD code was shown to have sufficient capability to evaluate the heat exchanger thermal-hydraulic performance.« less

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilkey, Lindsay

This milestone presents a demonstration of the High-to-Low (Hi2Lo) process in the VVI focus area. Validation and additional calculations with the commercial computational fluid dynamics code, STAR-CCM+, were performed using a 5x5 fuel assembly with non-mixing geometry and spacer grids. This geometry was based on the benchmark experiment provided by Westinghouse. Results from the simulations were compared to existing experimental data and to the subchannel thermal-hydraulics code COBRA-TF (CTF). An uncertainty quantification (UQ) process was developed for the STAR-CCM+ model and results of the STAR UQ were communicated to CTF. Results from STAR-CCM+ simulations were used as experimental design pointsmore » in CTF to calibrate the mixing parameter β and compared to results obtained using experimental data points. This demonstrated that CTF’s β parameter can be calibrated to match existing experimental data more closely. The Hi2Lo process for the STAR-CCM+/CTF code coupling was documented in this milestone and closely linked L3:VVI.H2LP15.01 milestone report.« less

1DTempPro: analyzing temperature profiles for groundwater/surface-water exchange

USGS Publications Warehouse

Voytek, Emily B.; Drenkelfuss, Anja; Day-Lewis, Frederick D.; Healy, Richard; Lane, John W.; Werkema, Dale D.

2014-01-01

A new computer program, 1DTempPro, is presented for the analysis of vertical one-dimensional (1D) temperature profiles under saturated flow conditions. 1DTempPro is a graphical user interface to the U.S. Geological Survey code Variably Saturated 2-Dimensional Heat Transport (VS2DH), which numerically solves the flow and heat-transport equations. Pre- and postprocessor features allow the user to calibrate VS2DH models to estimate vertical groundwater/surface-water exchange and also hydraulic conductivity for cases where hydraulic head is known.

Integration of fracturing dynamics and pressure transient analysis for hydraulic fracture evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arihara, N.; Abbaszadeh, M.; Wright, C.A.

This paper presents pre- and post-fracture pressure transient analysis, combined with net fracture pressure interpretation, for a well in a naturally fractured geothermal reservoir. Integrated analysis was performed to achieve a consistent interpretation of the created fracture geometry, propagation, conductivity, shrinkage, reservoir flow behavior, and formation permeability characteristics. The interpreted data includes two-rate pre-frac injection tests, step-rate injection tests, a series of pressure falloff tests, and the net fracturing pressure from a massive fracture treatment. Pressure transient analyses were performed utilizing advanced well test interpretation techniques and a thermal reservoir simulator with fracture propagation option. Hydraulic fracture propagation analysis wasmore » also performed Milt a generalized 3-D dynamic fracture growth model simulator. Three major conclusions resulted from the combined analysis: (1) that an increasing number of hydraulic fractures were being simultaneously propagated during the fracture treatment. (2) that the reservoir behaved as a composite reservoir Keith the outer region permeability being greater than the permeability of the region immediately surrounding the wellbore, and (3) that the created fractures extended into the outer region during the fracture treatment but retreated to the inner region several days after stimulation had ceased. These conclusions were apparent from independent pressure transient analysis and from independent hydraulic fracture propagation analysis. Integrated interpretation, however, increased the confidence in these conclusions and greatly aided the quantification of the created hydraulic fracture geometry and characterization of the reservoir permeability.« less

Modeling Thermal Pressurization Around Shallow Dikes Using Temperature-Dependent Hydraulic Properties: Implications for Deformation Around Intrusions

NASA Astrophysics Data System (ADS)

Townsend, Meredith R.

2018-01-01

Pressurization and flow of groundwater around igneous intrusions depend in part on the hydraulic diffusivity of the host rocks and processes that enhance diffusivity, such as fracturing, or decrease diffusivity, such as mineral precipitation during chemical alteration. Characterizing and quantifying the coupled effects of alteration, pore pressurization, and deformation have significant implications for deformation around intrusions, geothermal energy, contact metamorphism, and heat transfer at mid-ocean ridges. Fractures around dikes at Ship Rock, New Mexico, indicate that pore pressures in the host rocks exceeded hydrostatic conditions by at least 15 MPa following dike emplacement. Hydraulic measurements and petrographic analysis indicate that mineral precipitation clogged the pores of the host rock, reducing porosity from 0.25 to <0.10 and reducing permeability by 5 orders of magnitude. Field data from Ship Rock are used to motivate and constrain numerical models for thermal pore fluid pressurization adjacent to a meter-scale dike, using temperature-dependent hydraulic properties in the host rock as a proxy for porosity loss by mineral precipitation during chemical alteration. Reduction in permeability by chemical alteration has a negligible effect on pressurization. However, reduction in porosity by mineral precipitation increases fluid pressure by constricting pore volume and is identified as a potentially significant source of pressure. A scaling relationship is derived to determine when porosity loss becomes important; if permeability is low enough, pressurization by porosity loss outweighs pressurization by thermal expansion of fluids.

Dynamic Stability of the Rate, State, Temperature, and Pore Pressure Friction Model at a Rock Interface

NASA Astrophysics Data System (ADS)

Sinha, Nitish; Singh, Arun K.; Singh, Trilok N.

2018-05-01

In this article, we study numerically the dynamic stability of the rate, state, temperature, and pore pressure friction (RSTPF) model at a rock interface using standard spring-mass sliding system. This particular friction model is a basically modified form of the previously studied friction model namely the rate, state, and temperature friction (RSTF). The RSTPF takes into account the role of thermal pressurization including dilatancy and permeability of the pore fluid due to shear heating at the slip interface. The linear stability analysis shows that the critical stiffness, at which the sliding becomes stable to unstable or vice versa, increases with the coefficient of thermal pressurization. Critical stiffness, on the other hand, remains constant for small values of either dilatancy factor or hydraulic diffusivity, but the same decreases as their values are increased further from dilatancy factor (˜ 10^{ - 4} ) and hydraulic diffusivity (˜ 10^{ - 9} {m}2 {s}^{ - 1} ) . Moreover, steady-state friction is independent of the coefficient of thermal pressurization, hydraulic diffusivity, and dilatancy factor. The proposed model is also used for predicting time of failure of a creeping interface of a rock slope under the constant gravitational force. It is observed that time of failure decreases with increase in coefficient of thermal pressurization and hydraulic diffusivity, but the dilatancy factor delays the failure of the rock fault under the condition of heat accumulation at the creeping interface. Moreover, stiffness of the rock-mass also stabilizes the failure process of the interface as the strain energy due to the gravitational force accumulates in the rock-mass before it transfers to the sliding interface. Practical implications of the present study are also discussed.

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

NASA Astrophysics Data System (ADS)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Morlang, G.M.

1996-06-01

The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank throughmore » the flow visualization module to be radiographed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluorinert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system by remote operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the AP600 plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel/core model for comparison with the thermal-hydraulic codes.« less

An approach for coupled-code multiphysics core simulations from a common input

DOE PAGES

Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; ...

2014-12-10

This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which ismore » built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.« less

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

2011-07-01

In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of amore » power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

Integration of design, structural, thermal and optical analysis: And user's guide for structural-to-optical translator (PATCOD)

NASA Technical Reports Server (NTRS)

Amundsen, R. M.; Feldhaus, W. S.; Little, A. D.; Mitchum, M. V.

1995-01-01

Electronic integration of design and analysis processes was achieved and refined at Langley Research Center (LaRC) during the development of an optical bench for a laser-based aerospace experiment. Mechanical design has been integrated with thermal, structural and optical analyses. Electronic import of the model geometry eliminates the repetitive steps of geometry input to develop each analysis model, leading to faster and more accurate analyses. Guidelines for integrated model development are given. This integrated analysis process has been built around software that was already in use by designers and analysis at LaRC. The process as currently implemented used Pro/Engineer for design, Pro/Manufacturing for fabrication, PATRAN for solid modeling, NASTRAN for structural analysis, SINDA-85 and P/Thermal for thermal analysis, and Code V for optical analysis. Currently, the only analysis model to be built manually is the Code V model; all others can be imported for the Pro/E geometry. The translator from PATRAN results to Code V optical analysis (PATCOD) was developed and tested at LaRC. Directions for use of the translator or other models are given.

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure ofmore » the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.« less

Preliminary Thermal Modeling of HI-Storm 100S-218 Version B Storage Modules at Hope Creek Cuclear Power Station ISFSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuta, Judith M.; Adkins, Harold E.

2013-08-30

As part of the Used Fuel Disposition Campaign of the U. S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development, a consortium of national laboratories and industry is performing visual inspections and temperature measurements of selected storage modules at various locations around the United States. This report documents thermal analyses in in support of the inspections at the Hope Creek Nuclear Generating Station ISFSI. This site utilizes the HI-STORM100 vertical storage system developed by Holtec International. This is a vertical storage module design, and the thermal models are being developed using COBRA-SFS (Michener, et al.,more » 1987), a code developed by PNNL for thermal-hydraulic analyses of multi assembly spent fuel storage and transportation systems. This report describes the COBRA-SFS model in detail, and presents pre-inspection predictions of component temperatures and temperature distributions. The final report will include evaluation of inspection results, and if required, additional post-test calculations, with appropriate discussion of results.« less

A Thermal Management Systems Model for the NASA GTX RBCC Concept

NASA Technical Reports Server (NTRS)

Traci, Richard M.; Farr, John L., Jr.; Laganelli, Tony; Walker, James (Technical Monitor)

2002-01-01

The Vehicle Integrated Thermal Management Analysis Code (VITMAC) was further developed to aid the analysis, design, and optimization of propellant and thermal management concepts for advanced propulsion systems. The computational tool is based on engineering level principles and models. A graphical user interface (GUI) provides a simple and straightforward method to assess and evaluate multiple concepts before undertaking more rigorous analysis of candidate systems. The tool incorporates the Chemical Equilibrium and Applications (CEA) program and the RJPA code to permit heat transfer analysis of both rocket and air breathing propulsion systems. Key parts of the code have been validated with experimental data. The tool was specifically tailored to analyze rocket-based combined-cycle (RBCC) propulsion systems being considered for space transportation applications. This report describes the computational tool and its development and verification for NASA GTX RBCC propulsion system applications.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validatedmore » using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.« less

Target design optimization for an electron accelerator driven subcritical facility with circular and square beam profiles.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, M. Y. A; Sofu, T.; Zhong, Z.

2008-10-30

A subcritical facility driven by an electron accelerator is planned at the Kharkov Institute of Physics and Technology (KIPT) in Ukraine for medical isotope production, materials research, training, and education. The conceptual design of the facility is being pursued through collaborations between ANL and KIPT. As part of the design effort, the high-fidelity analyses of various target options are performed with formulations to reflect the realistic configuration and the three dimensional geometry of each design. This report summarizes the results of target design optimization studies for electron beams with two different beam profiles. The target design optimization is performed viamore » the sequential neutronic, thermal-hydraulic, and structural analyses for a comprehensive assessment of each configuration. First, a target CAD model is developed with proper emphasis on manufacturability to provide a basis for separate but consistent models for subsequent neutronic, thermal-hydraulic, and structural analyses. The optimizations are pursued for maximizing the neutron yield, streamlining the flow field to avoid hotspots, and minimizing the thermal stresses to increase the durability. In addition to general geometric modifications, the inlet/outlet channel configurations, target plate partitioning schemes, flow manipulations and rates, electron beam diameter/width options, and cladding material choices are included in the design optimizations. The electron beam interactions with the target assembly and the neutronic response of the subcritical facility are evaluated using the MCNPX code. the results for the electron beam energy deposition, neutron generation, and utilization in the subcritical pile are then used to characterize the axisymmetric heat generation profiles in the target assembly with explicit simulations of the beam tube, the coolant, the clad, and the target materials. Both tungsten and uranium are considered as target materials. Neutron spectra from tungsten and uranium are very similar allowing the use of either material in the subcritical assembly without changing its characteristics. However, the uranium target has a higher neutron yield, which increases the neutron flux of the subcritical assembly. Based on the considered dimensions and heat generation profiles, the commercial CFD software Star-CD is used for the thermal-hydraulic analysis of each target design to satisfy a set of thermal criteria, the most limiting of which being to maintain the water temperature 50 below the boiling point. It is found that the turbulence in the inlet channels dissipates quickly in narrow gaps between the target plates and, as a result, the heat transfer is limited by the laminar flow conditions. On average, 3-D CFD analyses of target assemblies agree well with 1-D calculations using RELAP (performed by KIPT). However, the recirculation and stagnation zones predicted with the CFD models prove the importance of a 3-D analysis to avoid the resulting hotspots. The calculated temperatures are subsequently used for the structural analysis of each target configuration to satisfy the other engineering design requirements. The thermo-structural calculations are performed mostly with NASTRAN and the results occasionally compared with the results from MARC. Both, NASTRAN and MARC are commercially available structural-mechanics analysis software. Although, a significant thermal gradient forms in target elements along the beam direction, the high thermal stresses are generally observed peripherally around the edge of thin target disks/plates. Due to its high thermal conductivity, temperatures and thermal stresses in tungsten target are estimated to be significantly lower than in uranium target. The deformations of the target disks/plates are found to be insignificant, which eliminate concerns for flow blockages in narrow coolant channels. Consistent with the specifications of the KIPT accelerator to be used in this facility, the electron beam power is 100-kW with electron energy in the range of 100 to 200 MeV. As expected, the 100 MeV electrons deposit their energy faster while the 200-MeV electrons spread their energy deposition further along the beam direction. However in that electron energy range, the energy deposition profiles near the beam window require very thin target plates/disks to limit the temperatures and thermal stresses.« less

ENVIRONMENTAL HYDRAULICS

EPA Science Inventory

The thermal, chemical, and biological quality of water in rivers, lakes, reservoirs, and near coastal areas is inseparable from a consideration of hydraulic engineering principles: therefore, the term environmental hydraulics. In this chapter we discuss the basic principles of w...

Thermal-hydraulic analysis of low activity fusion blanket designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fillo, J A; Powell, J; Yu, W S

1977-01-01

The heat transfer aspects of fusion blankets are considered where: (a) conduction and (b) boiling and condensation are the dominant heat transfer mechanisms. In some cases, unique heat transfer problems arise and additional heat transfer data and analyses may be required.

A comparison of two finite element models for the prediction of permafrost temperatures

DOT National Transportation Integrated Search

1986-01-01

In permafrost areas, the analysis of soil thermal and hydraulic regimes is complicated by phase change effects and by moisture migration. During freezing, mobile soil moisture is drawn toward the freezing front, increasing the available latent heat; ...

Scientific computations section monthly report, November 1993

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buckner, M.R.

1993-12-30

This progress report from the Savannah River Technology Center contains abstracts from papers from the computational modeling, applied statistics, applied physics, experimental thermal hydraulics, and packaging and transportation groups. Specific topics covered include: engineering modeling and process simulation, criticality methods and analysis, plutonium disposition.

Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

DOE PAGES

Salko, Robert K.; Schmidt, Rodney C.; Avramova, Maria N.

2014-11-23

This study describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis.

Ocean thermal gradient hydraulic power plant.

PubMed

Beck, E J

1975-07-25

Solar energy stored in the oceans may be used to generate power by exploiting ploiting thermal gradients. A proposed open-cycle system uses low-pressure steam to elevate vate water, which is then run through a hydraulic turbine to generate power. The device is analogous to an air lift pump.

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard; Bostelmann, F.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ling Zou; Hongbin Zhang; Jess Gehin

A coupled TH/Neutronics/CRUD framework, which is able to simulate the CRUD deposits impact on CIPS phenomenon, was described in this paper. This framework includes the coupling among three essential physics, thermal-hydraulics, CRUD and neutronics. The overall framework was implemented by using the CFD software STAR-CCM+, developing CRUD codes, and using the neutronics code DeCART. The coupling was implemented by exchanging data between softwares using intermediate exchange files. A typical 3 by 3 PWR fuel pin problem was solved under this framework. The problem was solved in a 12 months length period of time. Time-dependent solutions were provided, including CRUD depositsmore » inventory and their distributions on fuels, boron hideout amount inside CRUD deposits, as well as power shape changing over time. The results clearly showed the power shape suppression in regions where CRUD deposits exist, which is a strong indication of CIPS phenomenon.« less

Development of a New 47-Group Library for the CASL Neutronics Simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Williams, Mark L; Wiarda, Dorothea

The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACTmore » library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.« less

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todreas, N.E.; Cheng, S.K.; Basehore, K.

1984-08-01

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized.more » Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.« less

Preliminary design of a Primary Loop Pump Assembly (PLPA), using electromagnetic pumps

NASA Technical Reports Server (NTRS)

Moss, T. A.; Matlin, G.; Donelan, L.; Johnson, J. L.; Rowe, I.

1972-01-01

A preliminary design study of flight-type dc conduction-permanent magnetic, ac helical induction, and ac linear induction pumps for circulating 883 K (1130 F) NaK at 9.1 kg/sec (20 lb/sec) is described. Various electromagnetic pump geometrics are evaluated against hydraulic performance, and the effects of multiple windings and numbers of pumps per assembly on overall reliability were determined. The methods used in the electrical-hydraulic, stress, and thermal analysis are discussed, and the high temperature electrical materials selected for the application are listed.

CAVE: A computer code for two-dimensional transient heating analysis of conceptual thermal protection systems for hypersonic vehicles

NASA Technical Reports Server (NTRS)

Rathjen, K. A.

1977-01-01

A digital computer code CAVE (Conduction Analysis Via Eigenvalues), which finds application in the analysis of two dimensional transient heating of hypersonic vehicles is described. The CAVE is written in FORTRAN 4 and is operational on both IBM 360-67 and CDC 6600 computers. The method of solution is a hybrid analytical numerical technique that is inherently stable permitting large time steps even with the best of conductors having the finest of mesh size. The aerodynamic heating boundary conditions are calculated by the code based on the input flight trajectory or can optionally be calculated external to the code and then entered as input data. The code computes the network conduction and convection links, as well as capacitance values, given basic geometrical and mesh sizes, for four generations (leading edges, cooled panels, X-24C structure and slabs). Input and output formats are presented and explained. Sample problems are included. A brief summary of the hybrid analytical-numerical technique, which utilizes eigenvalues (thermal frequencies) and eigenvectors (thermal mode vectors) is given along with aerodynamic heating equations that have been incorporated in the code and flow charts.

Vadose zone dynamics governing snowmelt infiltration and groundwater recharge in a seasonally frozen, semi-arid landscape

NASA Astrophysics Data System (ADS)

Mohammed, A.; LeBlanc, F.; Cey, E. E.; Hayashi, M.

2016-12-01

Snowmelt infiltration and vadose zone fluxes in seasonally frozen soils are strongly affected by meteorological and soil moisture dynamics occurring during the preceding fall and winter, and complex processes controlling soil hydraulic and thermal regimes. In order to predict their effects on hydrologic processes such as run-off generation, groundwater recharge and plant-water availability in cold regions, an improved understanding of the mechanisms governing coupled water and heat fluxes in the unsaturated zone is needed. Field and laboratory studies were conducted to investigate snowmelt infiltration and groundwater recharge through partially frozen ground over a range of climate and soil conditions in the Canadian Prairies. Meteorological and subsurface field measurements at three sites were combined with laboratory infiltration experiments on frozen undisturbed soil-columns to provide insights into the hydraulic and thermal processes governing water movement. Analysis reveals that antecedent moisture content and thermal profiles both strongly affect subsurface dynamics during infiltration of snowmelt. Preferential flow is also a critical parameter, as both thermal and hydraulic responses were observed at depth prior to complete ground thaw in the field; as well as drainage outflow from the frozen soil column experiments under certain conditions. Results indicate that both diffuse (matrix) and preferential (macropore) flow play significant roles in the infiltration and redistribution of snowmelt water under frozen soil conditions, and shallow groundwater recharge. This study highlights the critical subsurface factors and processes that control infiltration and groundwater recharge in these seasonally frozen landscapes.

CASL L1 Milestone report : CASL.P4.01, sensitivity and uncertainty analysis for CIPS with VIPRE-W and BOA.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sung, Yixing; Adams, Brian M.; Secker, Jeffrey R.

2011-12-01

The CASL Level 1 Milestone CASL.P4.01, successfully completed in December 2011, aimed to 'conduct, using methodologies integrated into VERA, a detailed sensitivity analysis and uncertainty quantification of a crud-relevant problem with baseline VERA capabilities (ANC/VIPRE-W/BOA).' The VUQ focus area led this effort, in partnership with AMA, and with support from VRI. DAKOTA was coupled to existing VIPRE-W thermal-hydraulics and BOA crud/boron deposit simulations representing a pressurized water reactor (PWR) that previously experienced crud-induced power shift (CIPS). This work supports understanding of CIPS by exploring the sensitivity and uncertainty in BOA outputs with respect to uncertain operating and model parameters. Thismore » report summarizes work coupling the software tools, characterizing uncertainties, and analyzing the results of iterative sensitivity and uncertainty studies. These studies focused on sensitivity and uncertainty of CIPS indicators calculated by the current version of the BOA code used in the industry. Challenges with this kind of analysis are identified to inform follow-on research goals and VERA development targeting crud-related challenge problems.« less

Experiment on large scale plume interaction with a stratified gas environment resembling the thermal activity of a autocatalytic recombiner

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mignot, G.; Kapulla, R.; Paladino, D.

Computational Fluid Dynamics codes (CFD) are increasingly being used to simulate containment conditions after various transient accident scenarios. Consequently, the reliability of such codes must be tested against experimental data. Such validation experiments related to gas mixing and hydrogen transport within containment compartments addressing the effect of heat source are presented in this paper. The experiments were conducted in the large-scale thermal-hydraulics PANDA facility located at the Paul-Scherrer-Inst. (PSI) in Switzerland, in the frame of the OECD/SETH-2 project. A 10 kW electric heater simulating the thermal activity of the autocatalytic recombiner was activated at full power in a containment vesselmore » at the top of which a thick helium layer is initially present. The hot plume interacts with the bottom of the helium layer which is slowly eroded until complete break up at 1350 s. After final erosion of the layer a strong temperature and concentration gradient is maintained in the vessel below the heater inlet as well as in the adjacent vessel below the interconnecting pipe. A detailed characterization of the operating heater suggests the presence of cold gas ingress at the outlet that affects the flow in the chimney. This can be of concern if present in a real PAR unit. (authors)« less

Modeling and Simulation of the ITER First Wall/Blanket Primary Heat Transfer System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ying, Alice; Popov, Emilian L

2011-01-01

ITER inductive power operation is modeled and simulated using a thermal-hydraulics system code (RELAP5) integrated with a 3-D CFD (SC-Tetra) code. The Primary Heat Transfer System (PHTS) functions are predicted together with the main parameters operational ranges. The control algorithm strategy and derivation are summarized as well. The First Wall and Blanket modules are the primary components of PHTS, used to remove the major part of the thermal heat from the plasma. The modules represent a set of flow channels in solid metal structure that serve to absorb the radiation heat and nuclear heating from the fusion reactions and tomore » provide shield for the vacuum vessel. The blanket modules are water cooled. The cooling is forced convective with constant blanket inlet temperature and mass flow rate. Three independent water loops supply coolant to the three blanket sectors. The main equipment of each loop consists of a pump, a steam pressurizer and a heat exchanger. A major feature of ITER is the pulsed operation. The plasma does not burn continuously, but on intervals with large periods of no power between them. This specific feature causes design challenges to accommodate the thermal expansion of the coolant during the pulse period and requires active temperature control to maintain a constant blanket inlet temperature.« less

Thermal-hydraulic analysis capabilities and methods development at NYPA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1987-01-01

The operation of a nuclear power plant must be regularly supported by various thermal-hydraulic (T/H) analyses that may include final safety analysis report (FSAR) design basis calculations and licensing evaluations and conservative and best-estimate analyses. The development of in-house T/H capabilities provides the following advantages: (a) it leads to a better understanding of the plant design basis and operating characteristics; (b) methods developed can be used to optimize plant operations and enhance plant safety; (c) such a capability can be used for design reviews, checking vendor calculations, and evaluating proposed plant modifications; and (d) in-house capability reduces the cost ofmore » analysis. This paper gives an overview of the T/H capabilities and current methods development activity within the engineering department of the New York Power Authority (NYPA) and will focus specifically on reactor coolant system (RCS) transients and plant dynamic response for non-loss-of-coolant accident events. This paper describes NYPA experience in performing T/H analyses in support of pressurized water reactor plant operation.« less

Development of Safety Analysis Code System of Beam Transport and Core for Accelerator Driven System

NASA Astrophysics Data System (ADS)

Aizawa, Naoto; Iwasaki, Tomohiko

2014-06-01

Safety analysis code system of beam transport and core for accelerator driven system (ADS) is developed for the analyses of beam transients such as the change of the shape and position of incident beam. The code system consists of the beam transport analysis part and the core analysis part. TRACE 3-D is employed in the beam transport analysis part, and the shape and incident position of beam at the target are calculated. In the core analysis part, the neutronics, thermo-hydraulics and cladding failure analyses are performed by the use of ADS dynamic calculation code ADSE on the basis of the external source database calculated by PHITS and the cross section database calculated by SRAC, and the programs of the cladding failure analysis for thermoelastic and creep. By the use of the code system, beam transient analyses are performed for the ADS proposed by Japan Atomic Energy Agency. As a result, the rapid increase of the cladding temperature happens and the plastic deformation is caused in several seconds. In addition, the cladding is evaluated to be failed by creep within a hundred seconds. These results have shown that the beam transients have caused a cladding failure.

78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-05

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice... Reactor Fuels will hold a meeting on February 20, 2013, Room T-2B1, 11545 Rockville Pike, Rockville...

VERA 3.6 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L.; Kochunas, Brendan; Adams, Brian M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms.

Implementationof a modular software system for multiphysical processes in porous media

NASA Astrophysics Data System (ADS)

Naumov, Dmitri; Watanabe, Norihiro; Bilke, Lars; Fischer, Thomas; Lehmann, Christoph; Rink, Karsten; Walther, Marc; Wang, Wenqing; Kolditz, Olaf

2016-04-01

Subsurface georeservoirs are a candidate technology for large scale energy storage required as part of the transition to renewable energy sources. The increased use of the subsurface results in competing interests and possible impacts on protected entities. To optimize and plan the use of the subsurface in large scale scenario analyses,powerful numerical frameworks are required that aid process understanding and can capture the coupled thermal (T), hydraulic (H), mechanical (M), and chemical (C) processes with high computational efficiency. Due to having a multitude of different couplings between basic T, H, M, or C processes and the necessity to implement new numerical schemes the development focus has moved to software's modularity. The decreased coupling between the components results in two major advantages: easier addition of specialized processes and improvement of the code's testability and therefore its quality. The idea of modularization is implemented on several levels, in addition to library based separation of the previous code version, by using generalized algorithms available in the Standard Template Library and the Boost library, relying on efficient implementations of liner algebra solvers, using concepts when designing new types, and localization of frequently accessed data structures. This procedure shows certain benefits for a flexible high-performance framework applied to the analysis of multipurpose georeservoirs.

The Sixth Annual Thermal and Fluids Analysis Workshop

NASA Technical Reports Server (NTRS)

1995-01-01

The Sixth Annual Thermal and Fluids Analysis Workshop consisted of classes, vendor demonstrations, and paper sessions. The classes and vendor demonstrations provided participants with the information on widely used tools for thermal and fluids analysis. The paper sessions provided a forum for the exchange of information and ideas among thermal and fluids analysis. Paper topics included advances an uses of established thermal and fluids computer codes (such as SINDA and TRASYS) as well as unique modeling techniques and applications.

HT2DINV: A 2D forward and inverse code for steady-state and transient hydraulic tomography problems

NASA Astrophysics Data System (ADS)

Soueid Ahmed, A.; Jardani, A.; Revil, A.; Dupont, J. P.

2015-12-01

Hydraulic tomography is a technique used to characterize the spatial heterogeneities of storativity and transmissivity fields. The responses of an aquifer to a source of hydraulic stimulations are used to recover the features of the estimated fields using inverse techniques. We developed a 2D free source Matlab package for performing hydraulic tomography analysis in steady state and transient regimes. The package uses the finite elements method to solve the ground water flow equation for simple or complex geometries accounting for the anisotropy of the material properties. The inverse problem is based on implementing the geostatistical quasi-linear approach of Kitanidis combined with the adjoint-state method to compute the required sensitivity matrices. For undetermined inverse problems, the adjoint-state method provides a faster and more accurate approach for the evaluation of sensitivity matrices compared with the finite differences method. Our methodology is organized in a way that permits the end-user to activate parallel computing in order to reduce the computational burden. Three case studies are investigated demonstrating the robustness and efficiency of our approach for inverting hydraulic parameters.

Decay-ratio calculation in the frequency domain with the LAPUR code using 1D-kinetics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Munoz-Cobo, J. L.; Escriva, A.; Garcia, C.

This paper deals with the problem of computing the Decay Ratio in the frequency domain codes as the LAPUR code. First, it is explained how to calculate the feedback reactivity in the frequency domain using slab-geometry i.e. 1D kinetics, also we show how to perform the coupling of the 1D kinetics with the thermal-hydraulic part of the LAPUR code in order to obtain the reactivity feedback coefficients for the different channels. In addition, we show how to obtain the reactivity variation in the complex domain by solving the eigenvalue equation in the frequency domain and we compare this result withmore » the reactivity variation obtained in first order perturbation theory using the 1D neutron fluxes of the base case. Because LAPUR works in the linear regime, it is assumed that in general the perturbations are small. There is also a section devoted to the reactivity weighting factors used to couple the reactivity contribution from the different channels to the reactivity of the entire reactor core in point kinetics and 1D kinetics. Finally we analyze the effects of the different approaches on the DR value. (authors)« less

Simulation of Containment Atmosphere Mixing and Stratification Experiment in the ThAI Facility with a CFD Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babic, Miroslav; Kljenak, Ivo; Mavko, Borut

2006-07-01

The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. The main purpose of the proposed work was to assess the capabilities of the CFD code to reproduce the atmosphere structure with a three-dimensional model, coupled with condensation models proposed by the authors. A three-dimensional modelmore » of the ThAI vessel for the CFX4.4 code was developed. The flow in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also included. Calculated time-dependent variables together with temperature and volume fraction distributions at the end of different experiment phases are compared to experimental results. (authors)« less

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

2007-01-01

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

NASA Astrophysics Data System (ADS)

Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

2014-10-01

Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

Influence of spatial variability of hydraulic characteristics of soils on surface parameters obtained from remote sensing data in infrared and microwaves

NASA Technical Reports Server (NTRS)

Brunet, Y.; Vauclin, M.

1985-01-01

The correct interpretation of thermal and hydraulic soil parameters infrared from remotely sensed data (thermal infrared, microwaves) implies a good understanding of the causes of their temporal and spatial variability. Given this necessity, the sensitivity of the surface variables (temperature, moisture) to the spatial variability of hydraulic soil properties is tested with a numerical model of heat and mass transfer between bare soil and atmosphere. The spatial variability of hydraulic soil properties is taken into account in terms of the scaling factor. For a given soil, the knowledge of its frequency distribution allows a stochastic use of the model. The results are treated statistically, and the part of the variability of soil surface parameters due to that of soil hydraulic properties is evaluated quantitatively.

Arc-Heater Facility for Hot Hydrogen Exposure of Nuclear Thermal Rocket Materials

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Wang,Ten-See; Hickman, Robert; Panda, Binayak; Dobson, Chris; Osborne, Robin; Clifton, Scooter

2006-01-01

A hyper-thermal environment simulator is described for hot hydrogen exposure of nuclear thermal rocket material specimens and component development. This newly established testing capability uses a high-power, multi-gas, segmented arc-heater to produce high-temperature pressurized hydrogen flows representative of practical reactor core environments and is intended to serve. as a low cost test facility for the purpose of investigating and characterizing candidate fueUstructura1 materials and improving associated processing/fabrication techniques. Design and development efforts are thoroughly summarized, including thermal hydraulics analysis and simulation results, and facility operating characteristics are reported, as determined from a series of baseline performance mapping tests.

The Design of PSB-VVER Experiments Relevant to Accident Management

NASA Astrophysics Data System (ADS)

Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

Predicting Formation Damage in Aquifer Thermal Energy Storage Systems Utilizing a Coupled Hydraulic-Thermal-Chemical Reservoir Model

NASA Astrophysics Data System (ADS)

Müller, Daniel; Regenspurg, Simona; Milsch, Harald; Blöcher, Guido; Kranz, Stefan; Saadat, Ali

2014-05-01

In aquifer thermal energy storage (ATES) systems, large amounts of energy can be stored by injecting hot water into deep or intermediate aquifers. In a seasonal production-injection cycle, water is circulated through a system comprising the porous aquifer, a production well, a heat exchanger and an injection well. This process involves large temperature and pressure differences, which shift chemical equilibria and introduce or amplify mechanical processes. Rock-fluid interaction such as dissolution and precipitation or migration and deposition of fine particles will affect the hydraulic properties of the porous medium and may lead to irreversible formation damage. In consequence, these processes determine the long-term performance of the ATES system and need to be predicted to ensure the reliability of the system. However, high temperature and pressure gradients and dynamic feedback cycles pose challenges on predicting the influence of the relevant processes. Within this study, a reservoir model comprising a coupled hydraulic-thermal-chemical simulation was developed based on an ATES demonstration project located in the city of Berlin, Germany. The structural model was created with Petrel, based on data available from seismic cross-sections and wellbores. The reservoir simulation was realized by combining the capabilities of multiple simulation tools. For the reactive transport model, COMSOL Multiphysics (hydraulic-thermal) and PHREEQC (chemical) were combined using the novel interface COMSOL_PHREEQC, developed by Wissmeier & Barry (2011). It provides a MATLAB-based coupling interface between both programs. Compared to using COMSOL's built-in reactive transport simulator, PHREEQC additionally calculates adsorption and reaction kinetics and allows the selection of different activity coefficient models in the database. The presented simulation tool will be able to predict the most important aspects of hydraulic, thermal and chemical transport processes relevant to formation damage in ATES systems. We would like to present preliminary results of the structural reservoir model and the hydraulic-thermal-chemical coupling for the demonstration site. Literature: Wissmeier, L. and Barry, D.A., 2011. Simulation tool for variably saturated flow with comprehensive geochemical reactions in two- and three-dimensional domains. Environmental Modelling & Software 26, 210-218.

Thermal-hydraulic simulation of mercury target concepts for a pulsed spallation neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siman-Tov, M.; Wendel, M.; Haines, J.

1996-06-01

The Oak Ridge Spallation Neutron Source (ORSNS) is a high-power, accelerator-based pulsed spallation neutron source being designed by a multi-laboratory team led by Oak Ridge National Laboratory to achieve very high fluxes of neutrons for scientific experiments. The ORSNS is projected to have a 1 MW proton beam upgradable to 5 MW. About 60% of the beam power (1-5 MW, 17-83 kJ/pulse in 0.5 microsec at 60 cps) is deposited in the liquid metal (mercury) target having the dimensions of 65x30x10 cm (about 19.5 liter). Peak steady state power density is about 150 and 785 MW/m{sup 3} for 1 MWmore » and 5 MW beam respectively, whereas peak pulsed power density is as high as 5.2 and 26.1 GW/m{sup 3}, respectively. The peak pulse temperature rise rate is 14 million C/s (for 5 MW beam) whereas the total pulse temperature rise is only 7 C. In addition to thermal shock and materials compatibility, key feasibility issues for the target are related to its thermal-hydraulic performance. This includes proper flow distribution, flow reversals, possible {open_quotes}hot spots{close_quotes} and the challenge of mitigating the effects of thermal shock through possible injection of helium bubbles throughout the mercury volume or other concepts. The general computational fluid dynamics (CFD) code CFDS-FLOW3D was used to simulate the thermal and flow distribution in three preliminary concepts of the mercury target. Very initial CFD simulation of He bubbles injection demonstrates some potential for simulating behavior of He bubbles in flowing mercury. Much study and development will be required to be able to `predict`, even in a crude way, such a complex phenomena. Future direction in both design and R&D is outlined.« less

Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Singh, Gyanender P.; Gorton, Jacob

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperaturesmore » on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.« less

Method for calculating internal radiation and ventilation with the ADINAT heat-flow code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Butkovich, T.R.; Montan, D.N.

1980-04-01

One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation andmore » ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation.« less

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ishii, Mamoru

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results andmore » models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup power density on the flow instability. The experimental startup transient results showed the existence of three different flow instability mechanisms, i.e., flashing instability, condensation induced flow instability, and density wave oscillations. In addition, the void-reactivity feedback did not have significant effects on the flow instability during the startup transients for NMR-50. ii Several initial startup procedures with different power ramp rates were experimentally investigated to eliminate the flow instabilities observed from the startup transients. Particularly, the very slow startup transient and pressurized startup transient tests were performed and compared. It was found that the very slow startup transients by applying very small power density can eliminate the flashing oscillations in the single-phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. The initially pressurized startup procedure was tested to eliminate the flashing instability during the startup transients as well. The pressurized startup procedure included the initial pressurization, heat-up, and venting process. The startup transient tests showed that the pressurized startup procedure could eliminate the flow instability during the transition from single-phase flow to two-phase flow at low pressure conditions. The experimental results indicated that both startup procedures were applicable to the initial startup of NMR. However, the pressurized startup procedures might be preferred due to short operating hours required. In order to have a deeper understanding of natural circulation flow instability, the quasi-steady tests were performed using the test facility installed with preheater and subcooler. The effect of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback were investigated in the quasi-steady state tests. The experimental stability boundaries were determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR-type SMR. Two kinds of experiments, normal blowdown event and cold blowdown event, were experimentally investigated and compared with code predictions. The normal blowdown event was experimentally simulated since an initial condition where the pressure was lower than the designed pressure of the experiment facility, while the code prediction of blowdown started from the normal operation condition. Important thermal hydraulic parameters including reactor pressure vessel (RPV) pressure, containment pressure, local void fraction and temperature, pressure drop and natural circulation flow rate were measured and analyzed during the blowdown event. The pressure and water level transients are similar to the experimental results published by NuScale [51], which proves the capability of current loop in simulating the thermal hydraulic transient of real PWR-type SMR. During the 20000s blowdown experiment, water level in the core was always above the active fuel assemble during the experiment and proved the safety of natural circulation cooling and water recycling design of PWR-type SMR. Besides, pressure, temperature, and water level transient can be accurately predicted by RELAP5 code. However, the oscillations of natural circulation flow rate, water level and pressure drops were observed during the blowdown transients. This kind of flow oscillations are related to the water level and the location upper plenum, which is a path for coolant flow from chimney to steam generator and down comer. In order to investigate the transients start from the opening of ADS valve in both experimental and numerical way, the cold blow-down experiment is conducted. For the cold blowdown event, different from setting both reactor iv pressure vessel (RPV) and containment at high temperature and pressure, only RPV was heated close to the highest designed pressure and then open the ADS valve, same process was predicted using RELAP5 code. By doing cold blowdown experiment, the entire transients from the opening of ADS can be investigated by code and benchmarked with experimental data. Similar flow instability observed in the cold blowdown experiment. The comparison between code prediction and experiment data showed that the RELAP5 code can successfully predict the pressure void fraction and temperature transient during the cold blowdown event with limited error, but numerical instability exists in predicting natural circulation flow rate. Besides, the code is lack of capability in predicting the water level related flow instability observed in experiments.« less

Assessment of a hybrid finite element and finite volume code for turbulent incompressible flows

DOE PAGES

Xia, Yidong; Wang, Chuanjin; Luo, Hong; ...

2015-12-15

Hydra-TH is a hybrid finite-element/finite-volume incompressible/low-Mach flow simulation code based on the Hydra multiphysics toolkit being developed and used for thermal-hydraulics applications. In the present work, a suite of verification and validation (V&V) test problems for Hydra-TH was defined to meet the design requirements of the Consortium for Advanced Simulation of Light Water Reactors (CASL). The intent for this test problem suite is to provide baseline comparison data that demonstrates the performance of the Hydra-TH solution methods. The simulation problems vary in complexity from laminar to turbulent flows. A set of RANS and LES turbulence models were used in themore » simulation of four classical test problems. Numerical results obtained by Hydra-TH agreed well with either the available analytical solution or experimental data, indicating the verified and validated implementation of these turbulence models in Hydra-TH. Where possible, we have attempted some form of solution verification to identify sensitivities in the solution methods, and to suggest best practices when using the Hydra-TH code.« less

Assessment of a hybrid finite element and finite volume code for turbulent incompressible flows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xia, Yidong; Wang, Chuanjin; Luo, Hong

Hydra-TH is a hybrid finite-element/finite-volume incompressible/low-Mach flow simulation code based on the Hydra multiphysics toolkit being developed and used for thermal-hydraulics applications. In the present work, a suite of verification and validation (V&V) test problems for Hydra-TH was defined to meet the design requirements of the Consortium for Advanced Simulation of Light Water Reactors (CASL). The intent for this test problem suite is to provide baseline comparison data that demonstrates the performance of the Hydra-TH solution methods. The simulation problems vary in complexity from laminar to turbulent flows. A set of RANS and LES turbulence models were used in themore » simulation of four classical test problems. Numerical results obtained by Hydra-TH agreed well with either the available analytical solution or experimental data, indicating the verified and validated implementation of these turbulence models in Hydra-TH. Where possible, we have attempted some form of solution verification to identify sensitivities in the solution methods, and to suggest best practices when using the Hydra-TH code.« less

Modeling of Non-Homogeneous Containment Atmosphere in the ThAI Experimental Facility Using a CFD Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babic, Miroslav; Kljenak, Ivo; Mavko, Borut

2006-07-01

The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. The main purpose of the simulation was to reproduce the non-homogeneous temperature and species concentration distributions in the ThAI experimental facility. A three-dimensional model of the ThAI vessel for the CFX4.4 code was developed. The flowmore » in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also introduced. The calculated time-dependent variables together with temperature and concentration distributions at the end of experiment phases are compared to experimental results. (authors)« less

Experimental Investigation of Natural-Circulation Flow Behavior Under Low-Power/Low-Pressure Conditions in the Large-Scale PANDA Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Auban, Olivier; Paladino, Domenico; Zboray, Robert

2004-12-15

Twenty-five tests have been carried out in the large-scale thermal-hydraulic facility PANDA to investigate natural-circulation and stability behavior under low-pressure/low-power conditions, when void flashing might play an important role. This work, which extends the current experimental database to a large geometric scale, is of interest notably with regard to the start-up procedures in natural-circulation-cooled boiling water reactors. It should help the understanding of the physical phenomena that may cause flow instability in such conditions and can be used for validation of thermal-hydraulics system codes. The tests were performed at a constant power, balanced by a specific condenser heat removal capacity.more » The test matrix allowed the reactor pressure vessel power and pressure to be varied, as well as other parameters influencing the natural-circulation flow. The power spectra of flow oscillations showed in a few tests a major and unique resonance peak, and decay ratios between 0.5 and 0.9 have been found. The remainder of the tests showed an even more pronounced stable behavior. A classification of the tests is presented according to the circulation modes (from single-phase to two-phase flow) that could be assumed and particularly to the importance and the localization of the flashing phenomenon.« less

Structural-Thermal-Optical-Performance (STOP) Analysis

NASA Technical Reports Server (NTRS)

Bolognese, Jeffrey; Irish, Sandra

2015-01-01

The presentation will be given at the 26th Annual Thermal Fluids Analysis Workshop (TFAWS 2015) hosted by the Goddard Spaceflight Center (GSFC) Thermal Engineering Branch (Code 545). A STOP analysis is a multidiscipline analysis, consisting of Structural, Thermal and Optical Performance Analyses, that is performed for all space flight instruments and satellites. This course will explain the different parts of performing this analysis. The student will learn how to effectively interact with each discipline in order to accurately obtain the system analysis results.

Definition of hydraulic stability of KVGM-100 hot-water boiler and minimum water flow rate

NASA Astrophysics Data System (ADS)

Belov, A. A.; Ozerov, A. N.; Usikov, N. V.; Shkondin, I. A.

2016-08-01

In domestic power engineering, the methods of quantitative and qualitative-quantitative adjusting the load of the heat supply systems are widely distributed; furthermore, during the greater part of the heating period, the actual discharge of network water is less than estimated values when changing to quantitative adjustment. Hence, the hydraulic circuits of hot-water boilers should ensure the water velocities, minimizing the scale formation and excluding the formation of stagnant zones. The results of the calculations of hot-water KVGM-100 boiler and minimum water flow rate for the basic and peak modes at the fulfillment of condition of the lack of surface boil are presented in the article. The minimal flow rates of water at its underheating to the saturation state and the thermal flows in the furnace chamber were defined. The boiler hydraulic calculation was performed using the "Hydraulic" program, and the analysis of permissible and actual velocities of the water movement in the pipes of the heating surfaces was carried out. Based on the thermal calculations of furnace chamber and thermal- hydraulic calculations of heating surfaces, the following conclusions were drawn: the minimum velocity of water movement (by condition of boiling surface) at lifting movement of environment increases from 0.64 to 0.79 m/s; it increases from 1.14 to 1.38 m/s at down movement of environmental; the minimum water flow rate by the boiler in the basic mode (by condition of the surface boiling) increased from 887 t/h at the load of 20% up to 1074 t/h at the load of 100%. The minimum flow rate is 1074 t/h at nominal load and is achieved at the pressure at the boiler outlet equal to 1.1 MPa; the minimum water flow rate by the boiler in the peak mode by condition of surface boiling increases from 1669 t/h at the load of 20% up to 2021 t/h at the load of 100%.

Heat and Groundwater Flow in the San Gabriel Mountains, California

NASA Astrophysics Data System (ADS)

Newman, A. A.; Becker, M.; Laton, W. R., Jr.

2017-12-01

Groundwater flow paths in mountainous terrain often vary widely in both time and space. Such systems remain difficult to characterize due to fracture-dominated flow paths, high topographic relief, and sparse hydrologic data. We develop a hydrogeologic conceptual model of the Western San Gabriel Mountains in Southern California based on geophysical, thermal, and hydraulic head data. Boreholes are located along the San Gabriel Fault Zone (SGFZ) and cover a wide range of elevations to capture the heterogeneity of the hydrogeologic system. Long term (2016-2017) monitoring of temperature and hydraulic head was carried out in four shallow (300-600m depth) boreholes within the study area using fiber-optic distributed temperature sensing (DTS). Borehole temperature profiles were used to assess the regional groundwater flow system and local flows in fractures intersecting the borehole. DTS temperature profiles were compared with available borehole geophysical logs and head measurements collected with grouted vibrating wire pressure transducers (VWPT). Spatial and temporal variations in borehole temperature profiles suggest that advective heat transfer due to fluid flow affected the subsurface thermal regime. Thermal evidence of groundwater recharge and/or discharge and flow through discrete fractures was found in all four boreholes. Analysis of temporal changes to the flow system in response to seasonal and drilling-induced hydraulic forcing was useful in reducing ambiguities in noisy datasets and estimating interborehole relationships. Acoustic televiewer logs indicate fractures were primarily concentrated in densely fractured intervals, and only a minor decrease of fracture density was observed with depth. Anomalously high hydraulic gradients across the SGFZ suggest that the feature is a potential barrier to lateral flow. However, transient thermal anomalies consistent with groundwater flow within the SGFZ indicate this feature may be a potential conduit to vertical flow. This study builds upon the limited hydrogeologic understanding of the region and demonstrates the value of DTS in characterization efforts.

The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

NASA Astrophysics Data System (ADS)

Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

2015-04-01

The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They range from simpler, purely thermal cases (benchmark T1) to more complex, coupled 2D TH cases (benchmarks TH1, TH2, and TH3). Some experimental cases conducted in cold room complement the validation approach. A web site hosted by LSCE (Laboratoire des Sciences du Climat et de l'Environnement) is an interaction platform for the participants and hosts the test cases database at the following address: https://wiki.lsce.ipsl.fr/interfrost. The results of the first stage of the benchmark exercise will be presented. We will mainly focus on the inter-comparison of participant results for the coupled cases (TH1, TH2 & TH3). Further perspectives of the exercise will also be presented. Extensions to more complex physical conditions (e.g. unsaturated conditions and geometrical deformations) are contemplated. In addition, 1D vertical cases of interest to the Climate Modeling community will be proposed. Keywords: Permafrost; Numerical modeling; River-soil interaction; Arctic systems; soil freeze-thaw

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James; Bayless, Paul; Strydom, Gerhard

2016-11-01

Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations.more » Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.« less

RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumptionmore » of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. Themore » assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

Analysis of the quench propagation along Nb3Sn Rutherford cables with the THELMA code. Part I: Geometric and thermal models

NASA Astrophysics Data System (ADS)

Manfreda, G.; Bellina, F.

2016-12-01

The paper describes the new lumped thermal model recently implemented in THELMA code for the coupled electromagnetic-thermal analysis of superconducting cables. A new geometrical model is also presented, which describes the Rutherford cables used for the accelerator magnets. A first validation of these models has been given by the analysis of the quench longitudinal propagation velocity in the Nb3Sn prototype coil SMC3, built and tested in the frame of the EUCARD project for the development of high field magnets for LHC machine. This paper shows in detail the models, while their application to the quench propagation analysis is presented in a companion paper.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Comparison of the PHISICS/RELAP5-3D ring and block model results for phase I of the OECD/NEA MHTGR-350 benchmark

DOE PAGES

Strydom, G.; Epiney, A. S.; Alfonsi, Andrea; ...

2015-12-02

The PHISICS code system has been under development at INL since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and modules performing criticality searches, fuel shuffling and generalized perturbation. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort the first phase of the OECD/NEA MHTGR-350 Benchmark has now been completed. The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detailmore » in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 Benchmark and presents the results of Exercises 2 and 3 defined for Phase I. Exercise 2 required the modelling of a stand-alone thermal fluids solution at End of Equilibrium Cycle for the Modular High Temperature Reactor (MHTGR). The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity that can be obtained by this “block” model is illustrated with comparison results on the temperature, power density and flux distributions. Furthermore, it is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher fidelity block model, and that the additional model development and run-time efforts are worth the gains obtained in the improved spatial temperature and flux distributions.« less

1DTempPro: analyzing temperature profiles for groundwater/surface-water exchange.

PubMed

Voytek, Emily B; Drenkelfuss, Anja; Day-Lewis, Frederick D; Healy, Richard; Lane, John W; Werkema, Dale

2014-01-01

A new computer program, 1DTempPro, is presented for the analysis of vertical one-dimensional (1D) temperature profiles under saturated flow conditions. 1DTempPro is a graphical user interface to the U.S. Geological Survey code Variably Saturated 2-Dimensional Heat Transport (VS2DH), which numerically solves the flow and heat-transport equations. Pre- and postprocessor features allow the user to calibrate VS2DH models to estimate vertical groundwater/surface-water exchange and also hydraulic conductivity for cases where hydraulic head is known. Published 2013. This article is a U.S. Government work and is in the public domain in the USA.

Estimation of changes in dynamic hydraulic force in a magnetically suspended centrifugal blood pump with transient computational fluid dynamics analysis.

PubMed

Masuzawa, Toru; Ohta, Akiko; Tanaka, Nobuatu; Qian, Yi; Tsukiya, Tomonori

2009-01-01

The effect of the hydraulic force on magnetically levitated (maglev) pumps should be studied carefully to improve the suspension performance and the reliability of the pumps. A maglev centrifugal pump, developed at Ibaraki University, was modeled with 926 376 hexahedral elements for computational fluid dynamics (CFD) analyses. The pump has a fully open six-vane impeller with a diameter of 72.5 mm. A self-bearing motor suspends the impeller in the radial direction. The maximum pressure head and flow rate were 250 mmHg and 14 l/min, respectively. First, a steady-state analysis was performed using commercial code STAR-CD to confirm the model's suitability by comparing the results with the real pump performance. Second, transient analysis was performed to estimate the hydraulic force on the levitated impeller. The impeller was rotated in steps of 1 degrees using a sliding mesh. The force around the impeller was integrated at every step. The transient analysis revealed that the direction of the radial force changed dynamically as the vane's position changed relative to the outlet port during one circulation, and the magnitude of this force was about 1 N. The current maglev pump has sufficient performance to counteract this hydraulic force. Transient CFD analysis is not only useful for observing dynamic flow conditions in a centrifugal pump but is also effective for obtaining information about the levitation dynamics of a maglev pump.

Power generation costs and ultimate thermal hydraulic power limits in hypothetical advanced designs with natural circulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duffey, R.B.; Rohatgi, U.S.

Maximum power limits for hypothetical designs of natural circulation plants can be described analytically. The thermal hydraulic design parameters are those which limit the flow, being the elevations, flow areas, and loss coefficients. WE have found some simple ``design`` equations for natural circulation flow to power ratio, and for the stability limit. The analysis of historical and available data for maximum capacity factor estimation shows 80% to be reasonable and achievable. The least cost is obtained by optimizing both hypothetical plant performance for a given output,a nd the plant layout and design. There is also scope to increase output andmore » reduce cost by considering design variations of primary and secondary pressure, and by optimizing component elevations and loss coefficients. The design limits for each are set by stability and maximum flow considerations, which deserve close and careful evaluation.« less

Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

NASA Astrophysics Data System (ADS)

Bersano, Andrea

With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be compared to the values predicted by the system code and differences will be discussed with the ultimate goal to qualify RELAP5-3D for the analysis of decay heat removal systems in natural circulation. The numerical data will be also used to understand the key parameters related to the heat transfer in natural circulation and to optimize the operation of the system.

Numerical Analysis of Coolant Flow and Heat Transfer in ITER Diagnostic First Wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khodak, A.; Loesser, G.; Zhai, Y.

2015-07-24

We performed numerical simulations of the ITER Diagnostic First Wall (DFW) using ANSYS workbench. During operation DFW will include solid main body as well as liquid coolant. Thus thermal and hydraulic analysis of the DFW was performed using conjugated heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously fluid dynamics analysis was performed only in the liquid part. This approach includes interface between solid and liquid part of the systemAnalysis was performed using ANSYS CFX software. CFX software allows solution of heat transfer equations in solid and liquid part, and solution ofmore » the flow equations in the liquid part. Coolant flow in the DFW was assumed turbulent and was resolved using Reynolds averaged Navier-Stokes equations with Shear Stress Transport turbulence model. Meshing was performed using CFX method available within ANSYS. The data cloud for thermal loading consisting of volumetric heating and surface heating was imported into CFX Volumetric heating source was generated using Attila software. Surface heating was obtained using radiation heat transfer analysis. Our results allowed us to identify areas of excessive heating. Proposals for cooling channel relocation were made. Additional suggestions were made to improve hydraulic performance of the cooling system.« less

Altered Landscapes and Groundwater Sustainability — Exploring Impacts with Induced Polarization, DC Resistivity, and Thermal Tracing

NASA Astrophysics Data System (ADS)

Eddy-Miller, C.; Caldwell, R.; Wheeler, J.; McCarthy, P.; Binley, A. M.; Constantz, J. E.; Stonestrom, D. A.

2009-12-01

Anthropogenically impacted landscapes constitute rising proportions of the Earth’s surface that are characterized by generally elevated nutrient and sediment loadings concurrent with increased consumptive water withdrawals. In recent years a growing number of hydraulically engineered riparian habitat restoration projects have attempted to ameliorate negative impacts of land use on groundwater-surface water systems resulting, e.g., from agricultural practices and urban development. Often the nature of groundwater-surface water interactions in pre- and minimally altered systems is poorly known, making it difficult to assess the impacts of land use and restoration projects on groundwater sustainability. Traditional assessments of surface water parameters (flow, temperature, dissolved oxygen, biotic composition, etc.) can be complemented by hydraulic and thermal measurements to better understand the important role played by groundwater-surface water interactions. Hydraulic and thermal measurements are usually limited to point samples, however, making non-invasive and spatially extensive geophysical characterizations an attractive additional tool. Groundwater-surface water interactions along the Smith River, a tributary to the Missouri River in Montana, and Fish Creek and Flat Creek, tributaries to the Snake River in Wyoming, are being examined using a combination of hydraulic measurements, thermal tracing, and electrical-property imaging. Ninety-two direct-current (DC) resistivity and induced polarization cross sections were obtained at stream transects covering a wide variety of hydrogeologic settings ranging from shallow bedrock to thick alluvial sequences, nature of groundwater-surface water interactions (always gaining, always losing, or seasonally varying) and anthropogenic impacts (minimal low-intensity agriculture to major landscape engineering, including channel reconstruction). DC resistivity and induced polarization delineated mutually distinct features related to hydraulic architecture. For example, induced polarization imaging resolved channel-edge muck deposits that are presumed to be sites of low hydraulic conductivity, chemical reduction, and metal accumulation. DC resistivity delineated bedrock-alluvium contacts and showed potential for tracking changes in salinization. While electrical properties cannot substitute for hydraulic and thermal data, the addition of relatively rapidly acquired, spatially extensive resistivity and induced polarization imaging offers synergistic opportunities for interpretive hydrologic investigations.

Three-dimensional thermal analysis of a high-level waste repository

DOE Office of Scientific and Technical Information (OSTI.GOV)

Altenbach, T.J.

1979-04-01

The analysis used the TRUMP computer code to evaluate the thermal fields for six repository scenarios that studied the effects of room ventilation, room backfill, and repository thermal diffusivity. The results for selected nodes are presented as plots showing the effect of temperature as a function of time. 15 figures, 6 tables.

FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickson, T.L.

1993-01-01

This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracturemore » Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness.« less

FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickson, T.L.

1993-04-01

This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracturemore » Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness.« less

Flow distribution analysis on the cooling tube network of ITER thermal shield

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nam, Kwanwoo; Chung, Wooho; Noh, Chang Hyun

2014-01-29

Thermal shield (TS) is to be installed between the vacuum vessel or the cryostat and the magnets in ITER tokamak to reduce the thermal radiation load to the magnets operating at 4.2K. The TS is cooled by pressurized helium gas at the inlet temperature of 80K. The cooling tube is welded on the TS panel surface and the composed flow network of the TS cooling tubes is complex. The flow rate in each panel should be matched to the thermal design value for effective radiation shielding. This paper presents one dimensional analysis on the flow distribution of cooling tube networkmore » for the ITER TS. The hydraulic cooling tube network is modeled by an electrical analogy. Only the cooling tube on the TS surface and its connecting pipe from the manifold are considered in the analysis model. Considering the frictional factor and the local loss in the cooling tube, the hydraulic resistance is expressed as a linear function with respect to mass flow rate. Sub-circuits in the TS are analyzed separately because each circuit is controlled by its own control valve independently. It is found that flow rates in some panels are insufficient compared with the design values. In order to improve the flow distribution, two kinds of design modifications are proposed. The first one is to connect the tubes of the adjacent panels. This will increase the resistance of the tube on the panel where the flow rate is excessive. The other design suggestion is that an orifice is installed at the exit of tube routing where the flow rate is to be reduced. The analysis for the design suggestions shows that the flow mal-distribution is improved significantly.« less

CFD Analysis and Design of Detailed Target Configurations for an Accelerator-Driven Subcritical System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kraus, Adam; Merzari, Elia; Sofu, Tanju

2016-08-01

High-fidelity analysis has been utilized in the design of beam target options for an accelerator driven subcritical system. Designs featuring stacks of plates with square cross section have been investigated for both tungsten and uranium target materials. The presented work includes the first thermal-hydraulic simulations of the full, detailed target geometry. The innovative target cooling manifold design features many regions with complex flow features, including 90 bends and merging jets, which necessitate three-dimensional fluid simulations. These were performed using the commercial computational fluid dynamics code STAR-CCM+. Conjugate heat transfer was modeled between the plates, cladding, manifold structure, and fluid. Steady-statemore » simulations were performed but lacked good residual convergence. Unsteady simulations were then performed, which converged well and demonstrated that flow instability existed in the lower portion of the manifold. It was established that the flow instability had little effect on the peak plate temperatures, which were well below the melting point. The estimated plate surface temperatures and target region pressure were shown to provide sufficient margin to subcooled boiling for standard operating conditions. This demonstrated the safety of both potential target configurations during normal operation.« less

The analysis of thermal network of district heating system from investor point of view

NASA Astrophysics Data System (ADS)

Takács, Ján; Rácz, Lukáš

2016-06-01

The hydraulics of a thermal network of a district heating system is a very important issue, to which not enough attention is often paid. In this paper the authors want to point out some of the important aspects of the design and operation of thermal networks in district heating systems. The design boundary conditions of a heat distribution network and the requirements on active pressure - circulation pump - influencing the operation costs of the centralized district heating system as a whole, are analyzed in detail. The heat generators and the heat exchange stations are designed according to the design heat loads after thermal insulation, and modern boiler units are installed in the heating plant.

The use of multiobjective calibration and regional sensitivity analysis in simulating hyporheic exchange

USGS Publications Warehouse

Naranjo, Ramon C.; Niswonger, Richard G.; Stone, Mark; Davis, Clinton; McKay, Alan

2012-01-01

We describe an approach for calibrating a two-dimensional (2-D) flow model of hyporheic exchange using observations of temperature and pressure to estimate hydraulic and thermal properties. A longitudinal 2-D heat and flow model was constructed for a riffle-pool sequence to simulate flow paths and flux rates for variable discharge conditions. A uniform random sampling approach was used to examine the solution space and identify optimal values at local and regional scales. We used a regional sensitivity analysis to examine the effects of parameter correlation and nonuniqueness commonly encountered in multidimensional modeling. The results from this study demonstrate the ability to estimate hydraulic and thermal parameters using measurements of temperature and pressure to simulate exchange and flow paths. Examination of the local parameter space provides the potential for refinement of zones that are used to represent sediment heterogeneity within the model. The results indicate vertical hydraulic conductivity was not identifiable solely using pressure observations; however, a distinct minimum was identified using temperature observations. The measured temperature and pressure and estimated vertical hydraulic conductivity values indicate the presence of a discontinuous low-permeability deposit that limits the vertical penetration of seepage beneath the riffle, whereas there is a much greater exchange where the low-permeability deposit is absent. Using both temperature and pressure to constrain the parameter estimation process provides the lowest overall root-mean-square error as compared to using solely temperature or pressure observations. This study demonstrates the benefits of combining continuous temperature and pressure for simulating hyporheic exchange and flow in a riffle-pool sequence. Copyright 2012 by the American Geophysical Union.

System diagnostics using qualitative analysis and component functional classification

DOEpatents

Reifman, J.; Wei, T.Y.C.

1993-11-23

A method for detecting and identifying faulty component candidates during off-normal operations of nuclear power plants involves the qualitative analysis of macroscopic imbalances in the conservation equations of mass, energy and momentum in thermal-hydraulic control volumes associated with one or more plant components and the functional classification of components. The qualitative analysis of mass and energy is performed through the associated equations of state, while imbalances in momentum are obtained by tracking mass flow rates which are incorporated into a first knowledge base. The plant components are functionally classified, according to their type, as sources or sinks of mass, energy and momentum, depending upon which of the three balance equations is most strongly affected by a faulty component which is incorporated into a second knowledge base. Information describing the connections among the components of the system forms a third knowledge base. The method is particularly adapted for use in a diagnostic expert system to detect and identify faulty component candidates in the presence of component failures and is not limited to use in a nuclear power plant, but may be used with virtually any type of thermal-hydraulic operating system. 5 figures.

System diagnostics using qualitative analysis and component functional classification

DOEpatents

Reifman, Jaques; Wei, Thomas Y. C.

1993-01-01

A method for detecting and identifying faulty component candidates during off-normal operations of nuclear power plants involves the qualitative analysis of macroscopic imbalances in the conservation equations of mass, energy and momentum in thermal-hydraulic control volumes associated with one or more plant components and the functional classification of components. The qualitative analysis of mass and energy is performed through the associated equations of state, while imbalances in momentum are obtained by tracking mass flow rates which are incorporated into a first knowledge base. The plant components are functionally classified, according to their type, as sources or sinks of mass, energy and momentum, depending upon which of the three balance equations is most strongly affected by a faulty component which is incorporated into a second knowledge base. Information describing the connections among the components of the system forms a third knowledge base. The method is particularly adapted for use in a diagnostic expert system to detect and identify faulty component candidates in the presence of component failures and is not limited to use in a nuclear power plant, but may be used with virtually any type of thermal-hydraulic operating system.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

The kinetics of aerosol particle formation and removal in NPP severe accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.

2016-06-08

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal–hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into themore » KUPOL-M thermal–hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.« less

One-dimensional analysis of the hydrodynamic and thermal characteristics of thin film flows including the hydraulic jump and rotation

NASA Technical Reports Server (NTRS)

Thomas, S.; Hankey, W.; Faghri, A.; Swanson, T.

1990-01-01

The flow of a thin liquid film with a free surface along a horizontal plane that emanates from a pressurized vessel is examined numerically. In one g, a hydraulic jump was predicted in both plane and radial flow, which could be forced away from the inlet by increasing the inlet Froude number or Reynolds number. In zero g, the hydraulic jump was not predicted. The effect of solid-body rotation for radial flow in one g was to 'wash out' the hydraulic jump and to decrease the film height on the disk. The liquid film heights under one g and zero g were equal under solid-body rotation because the effect of centrifugal force was much greater than that of the gravitational force. The heat transfer to a film on a rotating disk was predicted to be greater than that of a stationary disk because the liquid film is extremely thin and is moving with a very high velocity.

Radiant Energy Measurements from a Scaled Jet Engine Axisymmetric Exhaust Nozzle for a Baseline Code Validation Case

NASA Technical Reports Server (NTRS)

Baumeister, Joseph F.

1994-01-01

A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.

THERMAL-ENERGY STORAGE IN A DEEP SANDSTONE AQUIFER IN MINNESOTA: FIELD OBSERVATIONS AND THERMAL ENERGY-TRANSPORT MODELING.

USGS Publications Warehouse

Miller, R.T.

1986-01-01

A study of the feasibility of storing heated water in a deep sandstone aquifer in Minnesota is described. The aquifer consists of four hydraulic zones that are areally anisotropic and have average hydraulic conductivities that range from 0. 03 to 1. 2 meters per day. A preliminary axially symmetric, nonisothermal, isotropic, single-phase, radial-flow, thermal-energy-transport model was constructed to investigate the sensitivity of model simulation to various hydraulic and thermal properties of the aquifer. A three-dimensional flow and thermal-energy transport model was constructed to incorporate the areal anisotropy of the aquifer. Analytical solutions of equations describing areally anisotropic groundwater flow around a doublet-well system were used to specify model boundary conditions for simulation of heat injection. The entire heat-injection-testing period of approximately 400 days was simulated. Model-computed temperatures compared favorably with field-recorded temperatures, with differences of no more than plus or minus 8 degree C. For each test cycle, model-computed aquifer thermal efficiency, defined as total heat withdrawn divided by total heat injected, was within plus or minus 2% of the field-calculated values.

MANTLE: A finite element program for the thermal-mechanical analysis of mantle convection. A user's manual with examples

NASA Technical Reports Server (NTRS)

Thompson, E.

1979-01-01

A finite element computer code for the analysis of mantle convection is described. The coupled equations for creeping viscous flow and heat transfer can be solved for either a transient analysis or steady-state analysis. For transient analyses, either a control volume or a control mass approach can be used. Non-Newtonian fluids with viscosities which have thermal and spacial dependencies can be easily incorporated. All material parameters may be written as function statements by the user or simply specified as constants. A wide range of boundary conditions, both for the thermal analysis and the viscous flow analysis can be specified. For steady-state analyses, elastic strain rates can be included. Although this manual was specifically written for users interested in mantle convection, the code is equally well suited for analysis in a number of other areas including metal forming, glacial flows, and creep of rock and soil.

Proceedings of the U.S. Army Symposium on Gun Dynamics (5th) Held in Rensselaerville, New York on 23-25 September 1987

DTIC Science & Technology

1987-09-01

have shown that gun barrel heating, and hence thermal expansion , is both axially and circumferentially asymmetric. Circumferential, or cross-barrel...element code, which ended in the selection of ABAQUS . The code will perform static, dynamic, and thermal anal- ysis on a broad range of structures...analysis may be performed by a user supplied FORTRAN subroutine which is automatically linked to the code and supplements the stand- ard ABAQUS

Physical and Hydrological Meaning of the Spectral Information from Hydrodynamic Signals at Karst Springs

NASA Astrophysics Data System (ADS)

Dufoyer, A.; Lecoq, N.; Massei, N.; Marechal, J. C.

2017-12-01

Physics-based modeling of karst systems remains almost impossible without enough accurate information about the inner physical characteristics. Usually, the only available hydrodynamic information is the flow rate at the karst outlet. Numerous works in the past decades have used and proven the usefulness of time-series analysis and spectral techniques applied to spring flow, precipitations or even physico-chemical parameters, for interpreting karst hydrological functioning. However, identifying or interpreting the karst systems physical features that control statistical or spectral characteristics of spring flow variations is still challenging, not to say sometimes controversial. The main objective of this work is to determine how the statistical and spectral characteristics of the hydrodynamic signal at karst springs can be related to inner physical and hydraulic properties. In order to address this issue, we undertake an empirical approach based on the use of both distributed and physics-based models, and on synthetic systems responses. The first step of the research is to conduct a sensitivity analysis of time-series/spectral methods to karst hydraulic and physical properties. For this purpose, forward modeling of flow through several simple, constrained and synthetic cases in response to precipitations is undertaken. It allows us to quantify how the statistical and spectral characteristics of flow at the outlet are sensitive to changes (i) in conduit geometries, and (ii) in hydraulic parameters of the system (matrix/conduit exchange rate, matrix hydraulic conductivity and storativity). The flow differential equations resolved by MARTHE, a computer code developed by the BRGM, allows karst conduits modeling. From signal processing on simulated spring responses, we hope to determine if specific frequencies are always modified, thanks to Fourier series and multi-resolution analysis. We also hope to quantify which parameters are the most variable with auto-correlation analysis: first results seem to show higher variations due to conduit conductivity than the ones due to matrix/conduit exchange rate. Future steps will be using another computer code, based on double-continuum approach and allowing turbulent conduit flow, and modeling a natural system.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operatingmore » pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.« less

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, Joo S.; Diamond, David

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

Nuclear Energy Knowledge and Validation Center (NEKVaC) Needs Workshop Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans

2015-02-01

The Department of Energy (DOE) has made significant progress developing simulation tools to predict the behavior of nuclear systems with greater accuracy and of increasing our capability to predict the behavior of these systems outside of the standard range of applications. These analytical tools require a more complex array of validation tests to accurately simulate the physics and multiple length and time scales. Results from modern simulations will allow experiment designers to narrow the range of conditions needed to bound system behavior and to optimize the deployment of instrumentation to limit the breadth and cost of the campaign. Modern validation,more » verification and uncertainty quantification (VVUQ) techniques enable analysts to extract information from experiments in a systematic manner and provide the users with a quantified uncertainty estimate. Unfortunately, the capability to perform experiments that would enable taking full advantage of the formalisms of these modern codes has progressed relatively little (with some notable exceptions in fuels and thermal-hydraulics); the majority of the experimental data available today is the "historic" data accumulated over the last decades of nuclear systems R&D. A validated code-model is a tool for users. An unvalidated code-model is useful for code developers to gain understanding, publish research results, attract funding, etc. As nuclear analysis codes have become more sophisticated, so have the measurement and validation methods and the challenges that confront them. A successful yet cost-effective validation effort requires expertise possessed only by a few, resources possessed only by the well-capitalized (or a willing collective), and a clear, well-defined objective (validating a code that is developed to satisfy the need(s) of an actual user). To that end, the Idaho National Laboratory established the Nuclear Energy Knowledge and Validation Center to address the challenges of modern code validation and to manage the knowledge from past, current, and future experimental campaigns. By pulling together the best minds involved in code development, experiment design, and validation to establish and disseminate best practices and new techniques, the Nuclear Energy Knowledge and Validation Center (NEKVaC or the ‘Center’) will be a resource for industry, DOE Programs, and academia validation efforts.« less

Advanced nodal neutron diffusion method with space-dependent cross sections: ILLICO-VX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rajic, H.L.; Ougouag, A.M.

1987-01-01

Advanced transverse integrated nodal methods for neutron diffusion developed since the 1970s require that node- or assembly-homogenized cross sections be known. The underlying structural heterogeneity can be accurately accounted for in homogenization procedures by the use of heterogeneity or discontinuity factors. Other (milder) types of heterogeneity, burnup-induced or due to thermal-hydraulic feedback, can be resolved by explicitly accounting for the spatial variations of material properties. This can be done during the nodal computations via nonlinear iterations. The new method has been implemented in the code ILLICO-VX (ILLICO variable cross-section method). Numerous numerical tests were performed. As expected, the convergence ratemore » of ILLICO-VX is lower than that of ILLICO, requiring approx. 30% more outer iterations per k/sub eff/ computation. The methodology has also been implemented as the NOMAD-VX option of the NOMAD, multicycle, multigroup, two- and three-dimensional nodal diffusion depletion code. The burnup-induced heterogeneities (space dependence of cross sections) are calculated during the burnup steps.« less

Numerical Simulation of the Emergency Condenser of the SWR-1000

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krepper, Eckhard; Schaffrath, Andreas; Aszodi, Attila

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Juelich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode.more » The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA Core Simulator (VERA-CS) development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code being developed by the Idaho National Laboratory (INL) is the focus of ongoing work inmore » the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework, BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.« less

RERTR-13 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

PWR PRELIMINARY DESIGN FOR PL-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humphries, G. E.

1962-02-28

The pressurized water reactor preliminary design, the preferred design developed under Phase I of the PL-3 contract, is presented. Plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentatlon description, shielding description, auxiliary systems, power plant equipment, waste dispusal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information, and a training program outline are given. (auth)

Development of concepts for the management of thermal resources in urban areas - Assessment of transferability from the Basel (Switzerland) and Zaragoza (Spain) case studies

NASA Astrophysics Data System (ADS)

Epting, Jannis; García-Gil, Alejandro; Huggenberger, Peter; Vázquez-Suñe, Enric; Mueller, Matthias H.

2017-05-01

The shallow subsurface in urban areas is increasingly used by shallow geothermal energy systems as a renewable energy resource and as a cheap cooling medium, e.g. for building air conditioning. In combination with further anthropogenic activities, this results in altered thermal regimes in the subsurface and the so-called subsurface urban heat island effect. Successful thermal management of urban groundwater resources requires understanding the relative contributions of the different thermal parameters and boundary conditions that result in the "present thermal state" of individual urban groundwater bodies. To evaluate the "present thermal state" of urban groundwater bodies, good quality data are required to characterize the hydraulic and thermal aquifer parameters. This process also involved adequate monitoring systems which provide consistent subsurface temperature measurements and are the basis for parameterizing numerical heat-transport models. This study is based on previous work already published for two urban groundwater bodies in Basel (CH) and Zaragoza (ES), where comprehensive monitoring networks (hydraulics and temperature) as well as calibrated high-resolution numerical flow- and heat-transport models have been analyzed. The "present thermal state" and how it developed according to the different hydraulic and thermal boundary conditions is compared to a "potential natural state" in order to assess the anthropogenic thermal changes that have already occurred in the urban groundwater bodies we investigated. This comparison allows us to describe the various processes concerning groundwater flow and thermal regimes for the different urban settings. Furthermore, the results facilitate defining goals for specific aquifer regions, including future aquifer use and urbanization, as well as evaluating the thermal use potential for these regions. As one example for a more sustainable thermal use of subsurface water resources, we introduce the thermal management concept of the "relaxation factor", which is a first approach to overcome the present policy of "first come, first served". Remediation measures to regenerate overheated urban aquifers are also introduced. The transferability of the applied methods to other urban areas is discussed. It is shown that an appropriate selection of locations for monitoring hydraulic and thermal boundary conditions make it possible to implement representative interpretations of groundwater flow and thermal regimes as well as to set up high-resolution numerical flow- and heat-transport models. Those models are the basis for the sustainable management of thermal resources.

Flexible parallel implicit modelling of coupled thermal-hydraulic-mechanical processes in fractured rocks

NASA Astrophysics Data System (ADS)

Cacace, Mauro; Jacquey, Antoine B.

2017-09-01

Theory and numerical implementation describing groundwater flow and the transport of heat and solute mass in fully saturated fractured rocks with elasto-plastic mechanical feedbacks are developed. In our formulation, fractures are considered as being of lower dimension than the hosting deformable porous rock and we consider their hydraulic and mechanical apertures as scaling parameters to ensure continuous exchange of fluid mass and energy within the fracture-solid matrix system. The coupled system of equations is implemented in a new simulator code that makes use of a Galerkin finite-element technique. The code builds on a flexible, object-oriented numerical framework (MOOSE, Multiphysics Object Oriented Simulation Environment) which provides an extensive scalable parallel and implicit coupling to solve for the multiphysics problem. The governing equations of groundwater flow, heat and mass transport, and rock deformation are solved in a weak sense (either by classical Newton-Raphson or by free Jacobian inexact Newton-Krylow schemes) on an underlying unstructured mesh. Nonlinear feedbacks among the active processes are enforced by considering evolving fluid and rock properties depending on the thermo-hydro-mechanical state of the system and the local structure, i.e. degree of connectivity, of the fracture system. A suite of applications is presented to illustrate the flexibility and capability of the new simulator to address problems of increasing complexity and occurring at different spatial (from centimetres to tens of kilometres) and temporal scales (from minutes to hundreds of years).

Verification and Validation Strategy for LWRS Tools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carl M. Stoots; Richard R. Schultz; Hans D. Gougar

2012-09-01

One intension of the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to create advanced computational tools for safety assessment that enable more accurate representation of a nuclear power plant safety margin. These tools are to be used to study the unique issues posed by lifetime extension and relicensing of the existing operating fleet of nuclear power plants well beyond their first license extension period. The extent to which new computational models / codes such as RELAP-7 can be used for reactor licensing / relicensing activities depends mainly upon the thoroughness with which they have been verifiedmore » and validated (V&V). This document outlines the LWRS program strategy by which RELAP-7 code V&V planning is to be accomplished. From the perspective of developing and applying thermal-hydraulic and reactivity-specific models to reactor systems, the US Nuclear Regulatory Commission (NRC) Regulatory Guide 1.203 gives key guidance to numeric model developers and those tasked with the validation of numeric models. By creating Regulatory Guide 1.203 the NRC defined a framework for development, assessment, and approval of transient and accident analysis methods. As a result, this methodology is very relevant and is recommended as the path forward for RELAP-7 V&V. However, the unique issues posed by lifetime extension will require considerations in addition to those addressed in Regulatory Guide 1.203. Some of these include prioritization of which plants / designs should be studied first, coupling modern supporting experiments to the stringent needs of new high fidelity models / codes, and scaling of aging effects.« less

TOUGH-RBSN simulator for hydraulic fracture propagation within fractured media: Model validations against laboratory experiments

NASA Astrophysics Data System (ADS)

Kim, Kunhwi; Rutqvist, Jonny; Nakagawa, Seiji; Birkholzer, Jens

2017-11-01

This paper presents coupled hydro-mechanical modeling of hydraulic fracturing processes in complex fractured media using a discrete fracture network (DFN) approach. The individual physical processes in the fracture propagation are represented by separate program modules: the TOUGH2 code for multiphase flow and mass transport based on the finite volume approach; and the rigid-body-spring network (RBSN) model for mechanical and fracture-damage behavior, which are coupled with each other. Fractures are modeled as discrete features, of which the hydrological properties are evaluated from the fracture deformation and aperture change. The verification of the TOUGH-RBSN code is performed against a 2D analytical model for single hydraulic fracture propagation. Subsequently, modeling capabilities for hydraulic fracturing are demonstrated through simulations of laboratory experiments conducted on rock-analogue (soda-lime glass) samples containing a designed network of pre-existing fractures. Sensitivity analyses are also conducted by changing the modeling parameters, such as viscosity of injected fluid, strength of pre-existing fractures, and confining stress conditions. The hydraulic fracturing characteristics attributed to the modeling parameters are investigated through comparisons of the simulation results.

New Challenges in Computational Thermal Hydraulics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yadigaroglu, George; Lakehal, Djamel

New needs and opportunities drive the development of novel computational methods for the design and safety analysis of light water reactors (LWRs). Some new methods are likely to be three dimensional. Coupling is expected between system codes, computational fluid dynamics (CFD) modules, and cascades of computations at scales ranging from the macro- or system scale to the micro- or turbulence scales, with the various levels continuously exchanging information back and forth. The ISP-42/PANDA and the international SETH project provide opportunities for testing applications of single-phase CFD methods to LWR safety problems. Although industrial single-phase CFD applications are commonplace, computational multifluidmore » dynamics is still under development. However, first applications are appearing; the state of the art and its potential uses are discussed. The case study of condensation of steam/air mixtures injected from a downward-facing vent into a pool of water is a perfect illustration of a simulation cascade: At the top of the hierarchy of scales, system behavior can be modeled with a system code; at the central level, the volume-of-fluid method can be applied to predict large-scale bubbling behavior; at the bottom of the cascade, direct-contact condensation can be treated with direct numerical simulation, in which turbulent flow (in both the gas and the liquid), interfacial dynamics, and heat/mass transfer are directly simulated without resorting to models.« less

The Fourth Annual Thermal and Fluids Analysis Workshop

NASA Technical Reports Server (NTRS)

1992-01-01

The Fourth Annual Thermal and Fluids Analysis Workshop was held from August 17-21, 1992, at NASA Lewis Research Center. The workshop consisted of classes, vendor demonstrations, and paper sessions. The classes and vendor demonstrations provided participants with the information on widely used tools for thermal and fluids analysis. The paper sessions provided a forum for the exchange of information and ideas among thermal and fluids analysts. Paper topics included advances and uses of established thermal and fluids computer codes (such as SINDA and TRASYS) as well as unique modeling techniques and applications.

TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1996-12-31

The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient.

Stand-alone containment analysis of Phébus FPT tests with ASTEC and MELCOR codes: the FPT-2 test.

PubMed

Gonfiotti, Bruno; Paci, Sandro

2018-03-01

During the last 40 years, many studies have been carried out to investigate the different phenomena occurring during a Severe Accident (SA) in a Nuclear Power Plant (NPP). Such efforts have been supported by the execution of different experimental campaigns, and the integral Phébus FP tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR) fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus FP campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel (CV) have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FP) behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FP and transuranic elements from a debris bed and a pool of melted fuel.

Thermal-Structural Optimization of Integrated Cryogenic Propellant Tank Concepts for a Reusable Launch Vehicle

NASA Technical Reports Server (NTRS)

Johnson, Theodore F.; Waters, W. Allen; Singer, Thomas N.; Haftka, Raphael T.

2004-01-01

A next generation reusable launch vehicle (RLV) will require thermally efficient and light-weight cryogenic propellant tank structures. Since these tanks will be weight-critical, analytical tools must be developed to aid in sizing the thickness of insulation layers and structural geometry for optimal performance. Finite element method (FEM) models of the tank and insulation layers were created to analyze the thermal performance of the cryogenic insulation layer and thermal protection system (TPS) of the tanks. The thermal conditions of ground-hold and re-entry/soak-through for a typical RLV mission were used in the thermal sizing study. A general-purpose nonlinear FEM analysis code, capable of using temperature and pressure dependent material properties, was used as the thermal analysis code. Mechanical loads from ground handling and proof-pressure testing were used to size the structural geometry of an aluminum cryogenic tank wall. Nonlinear deterministic optimization and reliability optimization techniques were the analytical tools used to size the geometry of the isogrid stiffeners and thickness of the skin. The results from the sizing study indicate that a commercial FEM code can be used for thermal analyses to size the insulation thicknesses where the temperature and pressure were varied. The results from the structural sizing study show that using combined deterministic and reliability optimization techniques can obtain alternate and lighter designs than the designs obtained from deterministic optimization methods alone.

A thermal NO(x) prediction model - Scalar computation module for CFD codes with fluid and kinetic effects

NASA Technical Reports Server (NTRS)

Mcbeath, Giorgio; Ghorashi, Bahman; Chun, Kue

1993-01-01

A thermal NO(x) prediction model is developed to interface with a CFD, k-epsilon based code. A converged solution from the CFD code is the input to the postprocessing model for prediction of thermal NO(x). The model uses a decoupled analysis to estimate the equilibrium level of (NO(x))e which is the constant rate limit. This value is used to estimate the flame (NO(x)) and in turn predict the rate of formation at each node using a two-step Zeldovich mechanism. The rate is fixed on the NO(x) production rate plot by estimating the time to reach equilibrium by a differential analysis based on the reaction: O + N2 = NO + N. The rate is integrated in the nonequilibrium time space based on the residence time at each node in the computational domain. The sum of all nodal predictions yields the total NO(x) level.

Analytical Assessment of a Gross Leakage Event Within the International Space Station (ISS) Node 2 Internal Active Thermal Control System (IATCS)

NASA Technical Reports Server (NTRS)

Holt, James M.; Clanton, Stephen E.

1999-01-01

Results of the International Space Station (ISS) Node 2 Internal Active Thermal Control System (IATCS) gross leakage analysis are presented for evaluating total leakage flowrates and volume discharge caused by a gross leakage event (i.e. open boundary condition). A Systems Improved Numerical Differencing Analyzer and Fluid Integrator (SINDA/FLUINT) thermal hydraulic mathematical model (THMM) representing the Node 2 IATCS was developed to simulate system performance under steady-state nominal conditions as well as the transient flow effects resulting from an open line exposed to ambient. The objective of the analysis was to determine the adequacy of the leak detection software in limiting the quantity of fluid lost during a gross leakage event to within an acceptable level.

Analytical Assessment of a Gross Leakage Event Within the International Space Station (ISS) Node 2 Internal Active Thermal Control System (IATCS)

NASA Technical Reports Server (NTRS)

Holt, James M.; Clanton, Stephen E.

2001-01-01

Results of the International Space Station (ISS) Node 2 Internal Active Thermal Control System (IATCS) gross leakage analysis are presented for evaluating total leakage flow rates and volume discharge caused by a gross leakage event (i.e. open boundary condition). A Systems Improved Numerical Differencing Analyzer and Fluid Integrator (SINDA85/FLUINT) thermal hydraulic mathematical model (THMM) representing the Node 2 IATCS was developed to simulate system performance under steady-state nominal conditions as well as the transient flow effect resulting from an open line exposed to ambient. The objective of the analysis was to determine the adequacy of the leak detection software in limiting the quantity of fluid lost during a gross leakage event to within an acceptable level.

Tools for Designing and Analyzing Structures

NASA Technical Reports Server (NTRS)

Luz, Paul L.

2005-01-01

Structural Design and Analysis Toolset is a collection of approximately 26 Microsoft Excel spreadsheet programs, each of which performs calculations within a different subdiscipline of structural design and analysis. These programs present input and output data in user-friendly, menu-driven formats. Although these programs cannot solve complex cases like those treated by larger finite element codes, these programs do yield quick solutions to numerous common problems more rapidly than the finite element codes, thereby making it possible to quickly perform multiple preliminary analyses - e.g., to establish approximate limits prior to detailed analyses by the larger finite element codes. These programs perform different types of calculations, as follows: 1. determination of geometric properties for a variety of standard structural components; 2. analysis of static, vibrational, and thermal- gradient loads and deflections in certain structures (mostly beams and, in the case of thermal-gradients, mirrors); 3. kinetic energies of fans; 4. detailed analysis of stress and buckling in beams, plates, columns, and a variety of shell structures; and 5. temperature dependent properties of materials, including figures of merit that characterize strength, stiffness, and deformation response to thermal gradients

Simulation of a main steam line break with steam generator tube rupture using trace

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallardo, S.; Querol, A.; Verdu, G.

A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation ofmore » the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)« less

The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Guentay, S.; Andreani, M.

2012-07-01

During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). Themore » tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)« less

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruggles, A.E.; Morris, D.G.

The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are usedmore » to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.« less

Theoretical investigation of the thermal hydraulic behaviour of a slab-type liquid metal target

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dury, T.V.; Smith, B.L.

1996-06-01

The thermal hydraulics codes CFDS-FLOW3D and ASTEC have been used to simulate a slabtype design of ESS spallation target. This design is single-skinned, and of tapering form (in the beam direction), with rounded sides in a cross-section through a plane normal to the beam. The coolant fluid used is mercury, under forced circulation, with an inlet temperature of 180{degrees}C. The goal of these computer studies was to understand the behaviour of the coolant flow, and hence to arrive at a design which optimises the heat extraction for a given beam power - in the sense of: (1) minimising the peakmore » local fluid temperature within the target, (2) maintaining an acceptable temperature level and distribution over and through the target outer wall, (3) keeping the overall fluid pressure loss through the complete target to a minimum, (4) staying within the physical limits of overall size required, particularly in the region of primary spallation. Two- and three-dimensional models have been used, with different arrangements and design of internal baffles, and different coolant flow distributions at the target inlet. Nominal total inlet mass flow was 245 kg/s, and a heat deposition profile used which was based on the proton beam energy distribution. This gave a nominal total heat load of 3.23 MW - of which 8.2kW were deposited in the window steel.« less

New PDC cutters improve drilling efficiency

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mensa-Wilmot, G.

1997-10-27

New polycrystalline diamond compact (PDC) cutters increase penetration rates and cumulative footage through improved abrasion, impact, interface strength, thermal stability, and fatigue characteristics. Studies of formation characterization, vibration analysis, hydraulic layouts, and bit selection continue to improve and expand PDC bit applications. The paper discusses development philosophy, performance characteristics and requirements, Types A, B, and C cutters, and combinations.

SEAWAT-based simulation of axisymmetric heat transport.

PubMed

Vandenbohede, Alexander; Louwyck, Andy; Vlamynck, Nele

2014-01-01

Simulation of heat transport has its applications in geothermal exploitation of aquifers and the analysis of temperature dependent chemical reactions. Under homogeneous conditions and in the absence of a regional hydraulic gradient, groundwater flow and heat transport from or to a well exhibit radial symmetry, and governing equations are reduced by one dimension (1D) which increases computational efficiency importantly. Solute transport codes can simulate heat transport and input parameters may be modified such that the Cartesian geometry can handle radial flow. In this article, SEAWAT is evaluated as simulator for heat transport under radial flow conditions. The 1971, 1D analytical solution of Gelhar and Collins is used to compare axisymmetric transport with retardation (i.e., as a result of thermal equilibrium between fluid and solid) and a large diffusion (conduction). It is shown that an axisymmetric simulation compares well with a fully three dimensional (3D) simulation of an aquifer thermal energy storage systems. The influence of grid discretization, solver parameters, and advection solution is illustrated. Because of the high diffusion to simulate conduction, convergence criterion for heat transport must be set much smaller (10(-10) ) than for solute transport (10(-6) ). Grid discretization should be considered carefully, in particular the subdivision of the screen interval. On the other hand, different methods to calculate the pumping or injection rate distribution over different nodes of a multilayer well lead to small differences only. © 2013, National Ground Water Association.

Receiver subsystem analysis report (RADL Item 4-1). The 10-MWe solar thermal central-receiver pilot plant: Solar-facilities design integration

NASA Astrophysics Data System (ADS)

1982-04-01

The results of thermal hydraulic, design for the stress analyses which are required to demonstrate that the receiver design for the Barstow Solar Pilot Plant satisfies the general design and performance requirements during the plant's design life are presented. Recommendations are made for receiver operation. The analyses are limited to receiver subsystem major structural parts (primary tower, receiver unit core support structure), pressure parts (absorber panels, feedwater, condensate and steam piping/components, flash tank, and steam mainfold) and shielding.

Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Dufek, Jan

2014-06-01

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

NASA Astrophysics Data System (ADS)

Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

2012-10-01

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

STAR-CCM+ Verification and Validation Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pointer, William David

2016-09-30

The commercial Computational Fluid Dynamics (CFD) code STAR-CCM+ provides general purpose finite volume method solutions for fluid dynamics and energy transport. This document defines plans for verification and validation (V&V) of the base code and models implemented within the code by the Consortium for Advanced Simulation of Light water reactors (CASL). The software quality assurance activities described herein are port of the overall software life cycle defined in the CASL Software Quality Assurance (SQA) Plan [Sieger, 2015]. STAR-CCM+ serves as the principal foundation for development of an advanced predictive multi-phase boiling simulation capability within CASL. The CASL Thermal Hydraulics Methodsmore » (THM) team develops advanced closure models required to describe the subgrid-resolution behavior of secondary fluids or fluid phases in multiphase boiling flows within the Eulerian-Eulerian framework of the code. These include wall heat partitioning models that describe the formation of vapor on the surface and the forces the define bubble/droplet dynamic motion. The CASL models are implemented as user coding or field functions within the general framework of the code. This report defines procedures and requirements for V&V of the multi-phase CFD capability developed by CASL THM. Results of V&V evaluations will be documented in a separate STAR-CCM+ V&V assessment report. This report is expected to be a living document and will be updated as additional validation cases are identified and adopted as part of the CASL THM V&V suite.« less

SHARP pre-release v1.0 - Current Status and Documentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahadevan, Vijay S.; Rahaman, Ronald O.

The NEAMS Reactor Product Line effort aims to develop an integrated multiphysics simulation capability for the design and analysis of future generations of nuclear power plants. The Reactor Product Line code suite’s multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. In this report, building on a several previous report issued in September 2014, we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representativemore » fast sodium-cooled reactor core in preparation for a unified release of the toolkit. The work reported in the current document covers the software engineering aspects of managing the entire stack of components in the SHARP toolkit and the continuous integration efforts ongoing to prepare a release candidate for interested reactor analysis users. Here we report on the continued integration effort of PROTEUS/Nek5000 and Diablo into the NEAMS framework and the software processes that enable users to utilize the capabilities without losing scientific productivity. Due to the complexity of the individual modules and their necessary/optional dependency library chain, we focus on the configuration and build aspects for the SHARP toolkit, which includes capability to autodownload dependencies and configure/install with optimal flags in an architecture-aware fashion. Such complexity is untenable without strong software engineering processes such as source management, source control, change reviews, unit tests, integration tests and continuous test suites. Details on these processes are provided in the report as a building step for a SHARP user guide that will accompany the first release, expected by Mar 2016.« less

Nuclear Analysis

NASA Technical Reports Server (NTRS)

Clement, J. D.; Kirby, K. D.

1973-01-01

Exploratory calculations were performed for several gas core breeder reactor configurations. The computational method involved the use of the MACH-1 one dimensional diffusion theory code and the THERMOS integral transport theory code for thermal cross sections. Computations were performed to analyze thermal breeder concepts and nonbreeder concepts. Analysis of breeders was restricted to the (U-233)-Th breeding cycle, and computations were performed to examine a range of parameters. These parameters include U-233 to hydrogen atom ratio in the gaseous cavity, carbon to thorium atom ratio in the breeding blanket, cavity size, and blanket size.

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szilard, Ronaldo Henriques

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toptan, Aysenur; Salko, Robert K; Avramova, Maria

Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes inmore » fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.« less

Ablative Thermal Response Analysis Using the Finite Element Method

NASA Technical Reports Server (NTRS)

Dec John A.; Braun, Robert D.

2009-01-01

A review of the classic techniques used to solve ablative thermal response problems is presented. The advantages and disadvantages of both the finite element and finite difference methods are described. As a first step in developing a three dimensional finite element based ablative thermal response capability, a one dimensional computer tool has been developed. The finite element method is used to discretize the governing differential equations and Galerkin's method of weighted residuals is used to derive the element equations. A code to code comparison between the current 1-D tool and the 1-D Fully Implicit Ablation and Thermal Response Program (FIAT) has been performed.

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Freels, James D; Bodey, Isaac T; Lowe, Kirk T

2010-09-01

The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Fluxmore » Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.« less

Phase Change Material Thermal Power Generator

NASA Technical Reports Server (NTRS)

Jones, Jack A. (Inventor); Chao, Yi (Inventor); Valdez, Thomas I. (Inventor)

2014-01-01

An energy producing device, for example a submersible vehicle for descending or ascending to different depths within water or ocean, is disclosed. The vehicle comprises a temperature-responsive material to which a hydraulic fluid is associated. A pressurized storage compartment stores the fluid as soon as the temperature-responsive material changes density. The storage compartment is connected with a hydraulic motor, and a valve allows fluid passage from the storage compartment to the hydraulic motor. An energy storage component, e.g. a battery, is connected with the hydraulic motor and is charged by the hydraulic motor when the hydraulic fluid passes through the hydraulic motor. Upon passage in the hydraulic motor, the fluid is stored in a further storage compartment and is then sent back to the area of the temperature-responsive material.

Phase change material thermal power generator

NASA Technical Reports Server (NTRS)

Jones, Jack A. (Inventor); Chao, Yi (Inventor); Valdez, Thomas I. (Inventor)

2011-01-01

An energy producing device, for example a submersible vehicle for descending or ascending to different depths within water or ocean, is disclosed. The vehicle comprises a temperature-responsive material to which a hydraulic fluid is associated. A pressurized storage compartment stores the fluid as soon as the temperature-responsive material changes density. The storage compartment is connected with a hydraulic motor, and a valve allows fluid passage from the storage compartment to the hydraulic motor. An energy storage component, e.g. a battery, is connected with the hydraulic motor and is charged by the hydraulic motor when the hydraulic fluid passes through the hydraulic motor. Upon passage in the hydraulic motor, the fluid is stored in a further storage compartment and is then sent back to the area of the temperature-responsive material.

Thermal neutron streaming effects and WIMS analysis of the Penn State subcritical graphite pile

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Zediak, C.S.; Jester, W.A.

1997-12-01

This analysis was performed on the Pennsylvania State University (PSU) subcritical reactor to find more accurate values for such nuclear parameters as the thermal fuel utilization factor, thermal diffusion length in the graphite, migration area, k{sub eff}, etc. The analysis involved using the Winfrith Integrated Multigroup Scheme (WIMS) code as well as various hand calculations to find and compare those parameters. The data found in this analysis will be used by future students in the Penn State laboratory courses.

LOFA analysis in helium and Pb-Li circuits of LLCB TBM by FE simulation

NASA Astrophysics Data System (ADS)

Chaudhuri, Paritosh; Ranjithkumar, S.; Sharma, Deepak; Danani, Chandan

2017-04-01

One of the main ITER objectives is to demonstrate the feasibility of the breeding blanket concepts that would lead to tritium self-sufficiency and the extraction of a high-grade heat for electricity production. India has developed the LLCB TBM to be tested in ITER for the validation of design concepts for tritium breeding blankets relevant DEMO and future power reactor. LLCB concept has the unique features of combination of both solid (lithium titanate as packed pebble bed) and liquid breeders (molten lead lithium). India specific IN-RAFMS is the structural material for TBM. The First Wall is actively cooled by high-pressure helium (He) gas [1]. It is important to validate the design of TBM to withstand various loads acting on it including accident analysis like LOCA, LOFA etc. Detailed thermal-hydraulic simulation studies including LOFA in helium and Pb-Li circuits of LLCB TBM have been performed using Finite Element using ANSYS. These analyses will provide important information about the temperature distribution in different materials used in TBM during steady state and transient condition. Thermal-hydraulic safety requirement has also been envisaged for the initiation the FPPS (Fusion Power Shutdown System) during LOFA. All these analysis will be presented in detail in this paper.

A Steady State and Quasi-Steady Interface Between the Generalized Fluid System Simulation Program and the SINDA/G Thermal Analysis Program

NASA Technical Reports Server (NTRS)

Schallhorn, Paul; Majumdar, Alok; Tiller, Bruce

2001-01-01

A general purpose, one dimensional fluid flow code is currently being interfaced with the thermal analysis program SINDA/G. The flow code, GFSSP, is capable of analyzing steady state and transient flow in a complex network. The flow code is capable of modeling several physical phenomena including compressibility effects, phase changes, body forces (such as gravity and centrifugal) and mixture thermodynamics for multiple species. The addition of GFSSP to SINDA/G provides a significant improvement in convective heat transfer modeling for SINDA/G. The interface development is conducted in multiple phases. This paper describes the first phase of the interface which allows for steady and quasisteady (unsteady solid, steady fluid) conjugate heat transfer modeling.

Summary of the thermal evaluation of LWBR (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lerner, S.; McWilliams, K.D.; Stout, J.W.

1980-03-01

This report describes the thermal evaluation of the core for the Shippingport Light Water Breeder Reactor. This core contains unique thermal-hydraulic features such as (1) close rod-to-rod proximity, (2) an open-lattice array of fuel rods with two different diameters and rod-to-rod spacings in the same flow region, (3) triplate orifices located at both the entrance and exit of fuel modules and (4) a hydraulically-balanced movable-fuel system coupled with (5) axial-and-radial fuel zoning for reactivity control. Performance studies used reactor thermal principles such as the hot-and-nominal channel concept and related nuclear/engineering design allowances. These were applied to models of three-dimensional roddedmore » arrays comprising the core fuel regions.« less

A method to combine hydrodynamics and constructive design in the optimization of the runner blades of Kaplan turbines

NASA Astrophysics Data System (ADS)

Miclosina, C. O.; Balint, D. I.; Campian, C. V.; Frunzaverde, D.; Ion, I.

2012-11-01

This paper deals with the optimization of the axial hydraulic turbines of Kaplan type. The optimization of the runner blade is presented systematically from two points of view: hydrodynamic and constructive. Combining these aspects in order to gain a safer operation when unsteady effects occur in the runner of the turbine is attempted. The design and optimization of the runner blade is performed with QTurbo3D software developed at the Center for Research in Hydraulics, Automation and Thermal Processes (CCHAPT) from "Eftimie Murgu" University of Resita, Romania. QTurbo3D software offers possibilities to design the meridian channel of hydraulic turbines design the blades and optimize the runner blade. 3D modeling and motion analysis of the runner blade operating mechanism are accomplished using SolidWorks software. The purpose of motion study is to obtain forces, torques or stresses in the runner blade operating mechanism, necessary to estimate its lifetime. This paper clearly states the importance of combining the hydrodynamics with the structural design in the optimization procedure of the runner of hydraulic turbines.

The National Spallation Neutron Source Target Station.

NASA Astrophysics Data System (ADS)

Gabriel, T. A.

1997-05-01

The technologies that are being utilized to design and build a state-of-the-art high powered (>= 1 MW), short pulsed (<= 1 μsec), and reliable spallation neutron source target station are discussed. The protons which directly and indirectly produce the neutrons will be obtained from a 1 GeV proton accelerator composed of an ion gun, rfq, linac, and storage ring. Many scientific and technical disciplines are required to produce a successful target station. These disciplines include engineering, remote handling, neutronics, materials, thermal hydraulics, shock analysis, etc. In the areas of engineering and remote handling special emphasis is being given to rapid and efficient assembly and disassembly of critical parts of the target station. In the neutronics area, emphasis is being given to neutron yield and pulse optimization from the moderators, and heating and activation rates throughout the station. Development of structural materials to withstand aggressive radiation environments and that are compatible with other materials is also an important area. Thermal hydraulics and shock analysis are being closely studied since large amounts of energy are being deposited in small volumes in relatively short time periods (< 1 μsec). These areas will be expanded upon in the paper.

Development of a thermal and structural analysis procedure for cooled radial turbines

NASA Technical Reports Server (NTRS)

Kumar, Ganesh N.; Deanna, Russell G.

1988-01-01

A procedure for computing the rotor temperature and stress distributions in a cooled radial turbine is considered. Existing codes for modeling the external mainstream flow and the internal cooling flow are used to compute boundary conditions for the heat transfer and stress analyses. An inviscid, quasi three-dimensional code computes the external free stream velocity. The external velocity is then used in a boundary layer analysis to compute the external heat transfer coefficients. Coolant temperatures are computed by a viscous one-dimensional internal flow code for the momentum and energy equation. These boundary conditions are input to a three-dimensional heat conduction code for calculation of rotor temperatures. The rotor stress distribution may be determined for the given thermal, pressure and centrifugal loading. The procedure is applied to a cooled radial turbine which will be tested at the NASA Lewis Research Center. Representative results from this case are included.

RERTR-8 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-8, was designed to test monolithic mini-fuel plates fabricated via hot isostatic pressing (HIP), the effect of molybdenum (Mo) content on the monolithic fuel behavior, and the efficiency of ternary additions to dispersion fuel particles on the interaction layer behavior at higher burnup. The following report summarizes the life of the RERTR-8 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

Mercury Thermal Hydraulic Loop (MTHL) Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Felde, David K.; Crye, Jason Michael; Wendel, Mark W.

2017-03-01

The Spallation Neutron Source (SNS) is a high-power linear accelerator built at Oak Ridge National Laboratory (ORNL) which incorporates the use of a flowing liquid mercury target. The Mercury Thermal Hydraulic Loop (MTHL) was constructed to investigate and verify the heat transfer characteristics of liquid mercury in a rectangular channel. This report provides a compilation of previously reported results from the water-cooled and electrically heated straight and curved test sections that simulate the geometry of the window cooling channel in the target nose region.

Utility of bromide and heat tracers for aquifer characterization affected by highly transient flow conditions

NASA Astrophysics Data System (ADS)

Ma, Rui; Zheng, Chunmiao; Zachara, John M.; Tonkin, Matthew

2012-08-01

A tracer test using both bromide and heat tracers conducted at the Integrated Field Research Challenge site in Hanford 300 Area (300A), Washington, provided an instrument for evaluating the utility of bromide and heat tracers for aquifer characterization. The bromide tracer data were critical to improving the calibration of the flow model complicated by the highly dynamic nature of the flow field. However, most bromide concentrations were obtained from fully screened observation wells, lacking depth-specific resolution for vertical characterization. On the other hand, depth-specific temperature data were relatively simple and inexpensive to acquire. However, temperature-driven fluid density effects influenced heat plume movement. Moreover, the temperature data contained "noise" caused by heating during fluid injection and sampling events. Using the hydraulic conductivity distribution obtained from the calibration of the bromide transport model, the temperature depth profiles and arrival times of temperature peaks simulated by the heat transport model were in reasonable agreement with observations. This suggested that heat can be used as a cost-effective proxy for solute tracers for calibration of the hydraulic conductivity distribution, especially in the vertical direction. However, a heat tracer test must be carefully designed and executed to minimize fluid density effects and sources of noise in temperature data. A sensitivity analysis also revealed that heat transport was most sensitive to hydraulic conductivity and porosity, less sensitive to thermal distribution factor, and least sensitive to thermal dispersion and heat conduction. This indicated that the hydraulic conductivity remains the primary calibration parameter for heat transport.

Utility of Bromide and Heat Tracers for Aquifer Characterization Affected by Highly Transient Flow Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ma, Rui; Zheng, Chunmiao; Zachara, John M.

A tracer test using both bromide and heat tracers conducted at the Integrated Field Research Challenge site in Hanford 300 Area (300A), Washington, provided an instrument for evaluating the utility of bromide and heat tracers for aquifer characterization. The bromide tracer data were critical to improving the calibration of the flow model complicated by the highly dynamic nature of the flow field. However, most bromide concentrations were obtained from fully screened observation wells, lacking depth-specific resolution for vertical characterization. On the other hand, depth-specific temperature data were relatively simple and inexpensive to acquire. However, temperature-driven fluid density effects influenced heatmore » plume movement. Moreover, the temperature data contained “noise” caused by heating during fluid injection and sampling events. Using the hydraulic conductivity distribution obtained from the calibration of the bromide transport model, the temperature depth profiles and arrival times of temperature peaks simulated by the heat transport model were in reasonable agreement with observations. This suggested that heat can be used as a cost-effective proxy for solute tracers for calibration of the hydraulic conductivity distribution, especially in the vertical direction. However, a heat tracer test must be carefully designed and executed to minimize fluid density effects and sources of noise in temperature data. A sensitivity analysis also revealed that heat transport was most sensitive to hydraulic conductivity and porosity, less sensitive to thermal distribution factor, and least sensitive to thermal dispersion and heat conduction. This indicated that the hydraulic conductivity remains the primary calibration parameter for heat transport.« less

Effect of Freeze-Thaw Cycles on Soil Nitrogen Reactive Transport in a Polygonal Arctic Tundra Ecosystem at Barrow AK Using 3-D Coupled ALM-PFLOTRAN

NASA Astrophysics Data System (ADS)

Yuan, F.; Wang, G.; Painter, S. L.; Tang, G.; Xu, X.; Kumar, J.; Bisht, G.; Hammond, G. E.; Mills, R. T.; Thornton, P. E.; Wullschleger, S. D.

2017-12-01

In Arctic tundra ecosystem soil freezing-thawing is one of dominant physical processes through which biogeochemical (e.g., carbon and nitrogen) cycles are tightly coupled. Besides hydraulic transport, freezing-thawing can cause pore water movement and aqueous species gradients, which are additional mechanisms for soil nitrogen (N) reactive-transport in Tundra ecosystem. In this study, we have fully coupled an in-development ESM(i.e., Advanced Climate Model for Energy, ACME)'s Land Model (ALM) aboveground processes with a state-of-the-art massively parallel 3-D subsurface thermal-hydrology and reactive transport code, PFLOTRAN. The resulting coupled ALM-PFLOTRAN model is a Land Surface Model (LSM) capable of resolving 3-D soil thermal-hydrological-biogeochemical cycles. This specific version of PFLOTRAN has incorporated CLM-CN Converging Trophic Cascade (CTC) model and a full and simple but robust soil N cycle. It includes absorption-desorption for soil NH4+ and gas dissolving-degasing process as well. It also implements thermal-hydrology mode codes with three newly-modified freezing-thawing algorithms which can greatly improve computing performance in regarding to numerical stiffness at freezing-point. Here we tested the model in fully 3-D coupled mode at the Next Generation Ecosystem Experiment-Arctic (NGEE-Arctic) field intensive study site at the Barrow Environmental Observatory (BEO), AK. The simulations show that: (1) synchronous coupling of soil thermal-hydrology and biogeochemistry in 3-D can greatly impact ecosystem dynamics across polygonal tundra landscape; and (2) freezing-thawing cycles can add more complexity to the system, resulting in greater mobility of soil N vertically and laterally, depending upon local micro-topography. As a preliminary experiment, the model is also implemented for Pan-Arctic region in 1-D column mode (i.e. no lateral connection), showing significant differences compared to stand-alone ALM. The developed ALM-PFLOTRAN coupling codes embeded within ESM will be used for Pan-Arctic regional evaluation of climate change-caused ecosystem responses and their feedbacks to climate system at various scales.

Analysis of a Distributed Pulse Power System Using a Circuit Analysis Code

DTIC Science & Technology

1979-06-01

dose rate was then integrated to give a number that could be compared with measure- ments made using thermal luminescent dosimeters ( TLD ’ s). Since...NM 8 7117 AND THE BDM CORPORATION, ALBUQUERQUE, NM 87106 Abstract A sophisticated computer code (SCEPTRE), used to analyze electronic circuits...computer code (SCEPTRE), used to analyze electronic circuits, was used to evaluate the performance of a large flash X-ray machine. This device was

Thermal Analysis on Plume Heating of the Main Engine on the Crew Exploration Vehicle Service Module

NASA Technical Reports Server (NTRS)

Wang, Xiao-Yen J.; Yuko, James R.

2007-01-01

The crew exploration vehicle (CEV) service module (SM) main engine plume heating is analyzed using multiple numerical tools. The chemical equilibrium compositions and applications (CEA) code is used to compute the flow field inside the engine nozzle. The plume expansion into ambient atmosphere is simulated using an axisymmetric space-time conservation element and solution element (CE/SE) Euler code, a computational fluid dynamics (CFD) software. The thermal analysis including both convection and radiation heat transfers from the hot gas inside the engine nozzle and gas radiation from the plume is performed using Thermal Desktop. Three SM configurations, Lockheed Martin (LM) designed 604, 605, and 606 configurations, are considered. Design of multilayer insulation (MLI) for the stowed solar arrays, which is subject to plume heating from the main engine, among the passive thermal control system (PTCS), are proposed and validated.

Interfacing comprehensive rotorcraft analysis with advanced aeromechanics and vortex wake models

NASA Astrophysics Data System (ADS)

Liu, Haiying

This dissertation describes three aspects of the comprehensive rotorcraft analysis. First, a physics-based methodology for the modeling of hydraulic devices within multibody-based comprehensive models of rotorcraft systems is developed. This newly proposed approach can predict the fully nonlinear behavior of hydraulic devices, and pressure levels in the hydraulic chambers are coupled with the dynamic response of the system. The proposed hydraulic device models are implemented in a multibody code and calibrated by comparing their predictions with test bench measurements for the UH-60 helicopter lead-lag damper. Predicted peak damping forces were found to be in good agreement with measurements, while the model did not predict the entire time history of damper force to the same level of accuracy. The proposed model evaluates relevant hydraulic quantities such as chamber pressures, orifice flow rates, and pressure relief valve displacements. This model could be used to design lead-lag dampers with desirable force and damping characteristics. The second part of this research is in the area of computational aeroelasticity, in which an interface between computational fluid dynamics (CFD) and computational structural dynamics (CSD) is established. This interface enables data exchange between CFD and CSD with the goal of achieving accurate airloads predictions. In this work, a loose coupling approach based on the delta-airloads method is developed in a finite-element method based multibody dynamics formulation, DYMORE. To validate this aerodynamic interface, a CFD code, OVERFLOW-2, is loosely coupled with a CSD program, DYMORE, to compute the airloads of different flight conditions for Sikorsky UH-60 aircraft. This loose coupling approach has good convergence characteristics. The predicted airloads are found to be in good agreement with the experimental data, although not for all flight conditions. In addition, the tight coupling interface between the CFD program, OVERFLOW-2, and the CSD program, DYMORE, is also established. The ability to accurately capture the wake structure around a helicopter rotor is crucial for rotorcraft performance analysis. In the third part of this thesis, a new representation of the wake vortex structure based on Non-Uniform Rational B-Spline (NURBS) curves and surfaces is proposed to develop an efficient model for prescribed and free wakes. NURBS curves and surfaces are able to represent complex shapes with remarkably little data. The proposed formulation has the potential to reduce the computational cost associated with the use of Helmholtz's law and the Biot-Savart law when calculating the induced flow field around the rotor. An efficient free-wake analysis will considerably decrease the computational cost of comprehensive rotorcraft analysis, making the approach more attractive to routine use in industrial settings.

The Fifth Annual Thermal and Fluids Analysis Workshop

NASA Technical Reports Server (NTRS)

1993-01-01

The Fifth Annual Thermal and Fluids Analysis Workshop was held at the Ohio Aerospace Institute, Brook Park, Ohio, cosponsored by NASA Lewis Research Center and the Ohio Aerospace Institute, 16-20 Aug. 1993. The workshop consisted of classes, vendor demonstrations, and paper sessions. The classes and vendor demonstrations provided participants with the information on widely used tools for thermal and fluid analysis. The paper sessions provided a forum for the exchange of information and ideas among thermal and fluids analysts. Paper topics included advances and uses of established thermal and fluids computer codes (such as SINDA and TRASYS) as well as unique modeling techniques and applications.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rouxelin, Pascal Nicolas; Strydom, Gerhard

Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented bymore » the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.« less

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khodak, A.; Zhai, Y.; Wang, W.

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of themore » coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.« less

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

DOE PAGES

Khodak, A.; Zhai, Y.; Wang, W.; ...

2017-06-19

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of themore » coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.« less

Thermal hydraulic feasibility assessment of the hot conditioning system and process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heard, F.J.

1996-10-10

The Spent Nuclear Fuel Project was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy`s Hanford Site in Richland, Washington. A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the proposed Hot Conditioning System and process for the Spent Nuclear Fuel Project. The analyses were performed using a series of thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the Hot Conditioning System. The subject efforts focus on independently investigating, quantifying,more » and establishing the governing heat production and removal mechanisms, flow distributions within the multi-canister overpack, and performing process simulations for various purge gases under consideration for the Hot Conditioning System, as well as obtaining preliminary results for comparison with and verification of other analyses, and providing technology- based recommendations for consideration and incorporation into the Hot Conditioning System design bases.« less

Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

NASA Astrophysics Data System (ADS)

Manfredini, A.; Mazzini, M.

2017-11-01

One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

Development of a thermal and structural analysis procedure for cooled radial turbines

NASA Technical Reports Server (NTRS)

Kumar, Ganesh N.; Deanna, Russell G.

1988-01-01

A procedure for computing the rotor temperature and stress distributions in a cooled radial turbine are considered. Existing codes for modeling the external mainstream flow and the internal cooling flow are used to compute boundary conditions for the heat transfer and stress analysis. The inviscid, quasi three dimensional code computes the external free stream velocity. The external velocity is then used in a boundary layer analysis to compute the external heat transfer coefficients. Coolant temperatures are computed by a viscous three dimensional internal flow cade for the momentum and energy equation. These boundary conditions are input to a three dimensional heat conduction code for the calculation of rotor temperatures. The rotor stress distribution may be determined for the given thermal, pressure and centrifugal loading. The procedure is applied to a cooled radial turbine which will be tested at the NASA Lewis Research Center. Representative results are given.

Development of concepts for the management of shallow geothermal resources in urban areas - Experience gained from the Basel and Zaragoza case studies

NASA Astrophysics Data System (ADS)

García-Gil, Alejandro; Epting, Jannis; Mueller, Matthias H.; Huggenberger, Peter; Vázquez-Suñé, Enric

2015-04-01

In urban areas the shallow subsurface often is used as a heat resource (shallow geothermal energy), i.e. for the installation and operation of a broad variety of geothermal systems. Increasingly, groundwater is used as a low-cost heat sink, e.g. for building acclimatization. Together with other shallow geothermal exploitation systems significantly increased groundwater temperatures have been observed in many urban areas (urban heat island effect). The experience obtained from two selected case study cities in Basel (CH) and Zaragoza (ES) has allowed developing concepts and methods for the management of thermal resources in urban areas. Both case study cities already have a comprehensive monitoring network operating (hydraulics and temperature) as well as calibrated high-resolution numerical groundwater flow and heat-transport models. The existing datasets and models have allowed to compile and compare the different hydraulic and thermal boundary conditions for both groundwater bodies, including: (1) River boundaries (River Rhine and Ebro), (2) Regional hydraulic and thermal settings, (3) Interaction with the atmosphere under consideration of urbanization and (4) Anthropogenic quantitative and thermal groundwater use. The potential natural states of the considered groundwater bodies also have been investigated for different urban settings and varying processes concerning groundwater flow and thermal regimes. Moreover, concepts for the management of thermal resources in urban areas and the transferability of the applied methods to other urban areas are discussed. The methods used provide an appropriate selection of parameters (spatiotemporal resolution) that have to be measured for representative interpretations of groundwater flow and thermal regimes of specific groundwater bodies. From the experience acquired from the case studies it is shown that understanding the variable influences of the specific geological and hydrogeological as well as hydraulic and thermal boundary conditions in urban settings is crucial. It also could be shown that good quality data are necessary to appropriately define and investigate thermal boundary conditions and the temperature development in urban systems. Groundwater temperatures in both investigated groundwater bodies are already over-heated and essentially impede further thermal groundwater use for cooling purposes. Current legislation approaches are not suitable to evaluate new concessions for thermal exploitation. Therefore, novel approaches for the assessment of new concessions which take into account the complex interaction of natural boundaries as well as existing shallow geothermal systems have to be developed.

Interfacing a General Purpose Fluid Network Flow Program with the SINDA/G Thermal Analysis Program

NASA Technical Reports Server (NTRS)

Schallhorn, Paul; Popok, Daniel

1999-01-01

A general purpose, one dimensional fluid flow code is currently being interfaced with the thermal analysis program Systems Improved Numerical Differencing Analyzer/Gaski (SINDA/G). The flow code, Generalized Fluid System Simulation Program (GFSSP), is capable of analyzing steady state and transient flow in a complex network. The flow code is capable of modeling several physical phenomena including compressibility effects, phase changes, body forces (such as gravity and centrifugal) and mixture thermodynamics for multiple species. The addition of GFSSP to SINDA/G provides a significant improvement in convective heat transfer modeling for SINDA/G. The interface development is conducted in multiple phases. This paper describes the first phase of the interface which allows for steady and quasi-steady (unsteady solid, steady fluid) conjugate heat transfer modeling.

Application of multiple-point geostatistics to simulate the effect of small-scale aquifer heterogeneity on the efficiency of aquifer thermal energy storage

NASA Astrophysics Data System (ADS)

Possemiers, Mathias; Huysmans, Marijke; Batelaan, Okke

2015-08-01

Adequate aquifer characterization and simulation using heat transport models are indispensible for determining the optimal design for aquifer thermal energy storage (ATES) systems and wells. Recent model studies indicate that meter-scale heterogeneities in the hydraulic conductivity field introduce a considerable uncertainty in the distribution of thermal energy around an ATES system and can lead to a reduction in the thermal recoverability. In a study site in Bierbeek, Belgium, the influence of centimeter-scale clay drapes on the efficiency of a doublet ATES system and the distribution of the thermal energy around the ATES wells are quantified. Multiple-point geostatistical simulation of edge properties is used to incorporate the clay drapes in the models. The results show that clay drapes have an influence both on the distribution of thermal energy in the subsurface and on the efficiency of the ATES system. The distribution of the thermal energy is determined by the strike of the clay drapes, with the major axis of anisotropy parallel to the clay drape strike. The clay drapes have a negative impact (3.3-3.6 %) on the energy output in the models without a hydraulic gradient. In the models with a hydraulic gradient, however, the presence of clay drapes has a positive influence (1.6-10.2 %) on the energy output of the ATES system. It is concluded that it is important to incorporate small-scale heterogeneities in heat transport models to get a better estimate on ATES efficiency and distribution of thermal energy.

Application of multiple-point geostatistics to simulate the effect of small scale aquifer heterogeneity on the efficiency of Aquifer Thermal Energy Storage (ATES)

NASA Astrophysics Data System (ADS)

Possemiers, Mathias; Huysmans, Marijke; Batelaan, Okke

2015-04-01

Adequate aquifer characterization and simulation using heat transport models are indispensible for determining the optimal design for Aquifer Thermal Energy Storage (ATES) systems and wells. Recent model studies indicate that meter scale heterogeneities in the hydraulic conductivity field introduce a considerable uncertainty in the distribution of thermal energy around an ATES system and can lead to a reduction in the thermal recoverability. In this paper, the influence of centimeter scale clay drapes on the efficiency of a doublet ATES system and the distribution of the thermal energy around the ATES wells are quantified. Multiple-point geostatistical simulation of edge properties is used to incorporate the clay drapes in the models. The results show that clay drapes have an influence both on the distribution of thermal energy in the subsurface and on the efficiency of the ATES system. The distribution of the thermal energy is determined by the strike of the clay drapes, with the major axis of anisotropy parallel to the clay drape strike. The clay drapes have a negative impact (3.3 - 3.6%) on the energy output in the models without a hydraulic gradient. In the models with a hydraulic gradient, however, the presence of clay drapes has a positive influence (1.6 - 10.2%) on the energy output of the ATES system. It is concluded that it is important to incorporate small scale heterogeneities in heat transport models to get a better estimate on ATES efficiency and distribution of thermal energy.

Development of Adaptive Model Refinement (AMoR) for Multiphysics and Multifidelity Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turinsky, Paul

This project investigated the development and utilization of Adaptive Model Refinement (AMoR) for nuclear systems simulation applications. AMoR refers to utilization of several models of physical phenomena which differ in prediction fidelity. If the highest fidelity model is judged to always provide or exceeded the desired fidelity, than if one can determine the difference in a Quantity of Interest (QoI) between the highest fidelity model and lower fidelity models, one could utilize the fidelity model that would just provide the magnitude of the QoI desired. Assuming lower fidelity models require less computational resources, in this manner computational efficiency can bemore » realized provided the QoI value can be accurately and efficiently evaluated. This work utilized Generalized Perturbation Theory (GPT) to evaluate the QoI, by convoluting the GPT solution with the residual of the highest fidelity model determined using the solution from lower fidelity models. Specifically, a reactor core neutronics problem and thermal-hydraulics problem were studied to develop and utilize AMoR. The highest fidelity neutronics model was based upon the 3D space-time, two-group, nodal diffusion equations as solved in the NESTLE computer code. Added to the NESTLE code was the ability to determine the time-dependent GPT neutron flux. The lower fidelity neutronics model was based upon the point kinetics equations along with utilization of a prolongation operator to determine the 3D space-time, two-group flux. The highest fidelity thermal-hydraulics model was based upon the space-time equations governing fluid flow in a closed channel around a heat generating fuel rod. The Homogenous Equilibrium Mixture (HEM) model was used for the fluid and Finite Difference Method was applied to both the coolant and fuel pin energy conservation equations. The lower fidelity thermal-hydraulic model was based upon the same equations as used for the highest fidelity model but now with coarse spatial meshing, corrected somewhat by employing effective fuel heat conduction values. The effectiveness of switching between the highest fidelity model and lower fidelity model as a function of time was assessed using the neutronics problem. Based upon work completed to date, one concludes that the time switching is effective in annealing out differences between the highest and lower fidelity solutions. The effectiveness of using a lower fidelity GPT solution, along with a prolongation operator, to estimate the QoI was also assessed. The utilization of a lower fidelity GPT solution was done in an attempt to avoid the high computational burden associated with solving for the highest fidelity GPT solution. Based upon work completed to date, one concludes that the lower fidelity adjoint solution is not sufficiently accurate with regard to estimating the QoI; however, a formulation has been revealed that may provide a path for addressing this shortcoming.« less

Thermal analysis of combinatorial solid geometry models using SINDA

NASA Technical Reports Server (NTRS)

Gerencser, Diane; Radke, George; Introne, Rob; Klosterman, John; Miklosovic, Dave

1993-01-01

Algorithms have been developed using Monte Carlo techniques to determine the thermal network parameters necessary to perform a finite difference analysis on Combinatorial Solid Geometry (CSG) models. Orbital and laser fluxes as well as internal heat generation are modeled to facilitate satellite modeling. The results of the thermal calculations are used to model the infrared (IR) images of targets and assess target vulnerability. Sample analyses and validation are presented which demonstrate code products.

LARGE-SCALE NATURAL GRADIENT TRACER TEST IN SAND AND GRAVEL, CAPE CODE, MASSACHUSETTS 3. HYDRAULIC CONDUCTI- VITY AND CALCULATED MACRODISPERSIVITIES

EPA Science Inventory

Hydraulic conductivity (K) variability in a sand and gravel aquifer on Cape Cod, Massachusetts, was measured and subsequently used in stochastic transport theories to estimate macrodispersivities. Nearly 1500 K measurements were obtained by borehole flowmeter tests ...

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

Molecular Tagging Velocimetry Development for In-situ Measurement in High-Temperature Test Facility

NASA Technical Reports Server (NTRS)

Andre, Matthieu A.; Bardet, Philippe M.; Burns, Ross A.; Danehy, Paul M.

2015-01-01

The High Temperature Test Facility, HTTF, at Oregon State University (OSU) is an integral-effect test facility designed to model the behavior of a Very High Temperature Gas Reactor (VHTR) during a Depressurized Conduction Cooldown (DCC) event. It also has the ability to conduct limited investigations into the progression of a Pressurized Conduction Cooldown (PCC) event in addition to phenomena occurring during normal operations. Both of these phenomena will be studied with in-situ velocity field measurements. Experimental measurements of velocity are critical to provide proper boundary conditions to validate CFD codes, as well as developing correlations for system level codes, such as RELAP5 (http://www4vip.inl.gov/relap5/). Such data will be the first acquired in the HTTF and will introduce a diagnostic with numerous other applications to the field of nuclear thermal hydraulics. A laser-based optical diagnostic under development at The George Washington University (GWU) is presented; the technique is demonstrated with velocity data obtained in ambient temperature air, and adaptation to high-pressure, high-temperature flow is discussed.

Condensation model for the ESBWR passive condensers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Revankar, S. T.; Zhou, W.; Wolf, B.

2012-07-01

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data frommore » separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE PAGES

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald; ...

2014-05-01

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Extending a CAD-Based Cartesian Mesh Generator for the Lattice Boltzmann Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantrell, J Nathan; Inclan, Eric J; Joshi, Abhijit S

2012-01-01

This paper describes the development of a custom preprocessor for the PaRAllel Thermal Hydraulics simulations using Advanced Mesoscopic methods (PRATHAM) code based on an open-source mesh generator, CartGen [1]. PRATHAM is a three-dimensional (3D) lattice Boltzmann method (LBM) based parallel flow simulation software currently under development at the Oak Ridge National Laboratory. The LBM algorithm in PRATHAM requires a uniform, coordinate system-aligned, non-body-fitted structured mesh for its computational domain. CartGen [1], which is a GNU-licensed open source code, already comes with some of the above needed functionalities. However, it needs to be further extended to fully support the LBM specificmore » preprocessing requirements. Therefore, CartGen is being modified to (i) be compiler independent while converting a neutral-format STL (Stereolithography) CAD geometry to a uniform structured Cartesian mesh, (ii) provide a mechanism for PRATHAM to import the mesh and identify the fluid/solid domains, and (iii) provide a mechanism to visually identify and tag the domain boundaries on which to apply different boundary conditions.« less

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

Design and Construction of Experiment for Direct Electron Irradiation of Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chemerisov, Sergey; Gromov, Roman; Makarashvili, Vakho

Argonne is assisting SHINE Medical Technologies in developing SHINE, a system for producing fission-product 99Mo using a D/T-accelerator to produce fission in a non-critical target solution of aqueous uranyl sulfate. We have developed an experimental setup for studying thermal-hydraulics and bubble formation in the uranyl sulfate solution to simulate conditions expected in the SHINE target solution during irradiation. A direct electron beam from the linac accelerator will be used to irradiate a 20 L solution (sector of the solution vessel). Because the solution will undergo radiolytic decomposition, we will be able to study bubble formation and dynamics and effects ofmore » convection and temperature on bubble behavior. These experiments will serve as a verification/ validation tool for the thermal-hydraulic model. Utilization of the direct electron beam for irradiation allows homogeneous heating of a large solution volume and simplifies observation of the bubble dynamics simultaneously with thermal-hydraulic data collection, which will complement data collected during operation of the miniSHINE experiment. Irradiation will be conducted using a 30-40 MeV electron beam from the high-power linac accelerator. The total electron-beam power will be 20 kW, which will yield a power density on the order of 1 kW/L. The solution volume will be cooled on the front and back surfaces and central tube to mimic the geometry of the proposed SHINE solution vessel. Also, multiple thermocouples will be inserted into the solution vessel to map thermal profiles. The experimental design is now complete, and installation and testing are in progress.« less

Experimental investigation of the hydraulic and heat-transfer properties of artificially fractured granite.

PubMed

Luo, Jin; Zhu, Yongqiang; Guo, Qinghai; Tan, Long; Zhuang, Yaqin; Liu, Mingliang; Zhang, Canhai; Xiang, Wei; Rohn, Joachim

2017-01-05

In this paper, the hydraulic and heat-transfer properties of two sets of artificially fractured granite samples are investigated. First, the morphological information is determined using 3D modelling technology. The area ratio is used to describe the roughness of the fracture surface. Second, the hydraulic properties of fractured granite are tested by exposing samples to different confining pressures and temperatures. The results show that the hydraulic properties of the fractures are affected mainly by the area ratio, with a larger area ratio producing a larger fracture aperture and higher hydraulic conductivity. Both the hydraulic apertureand the hydraulic conductivity decrease with an increase in the confining pressure. Furthermore, the fracture aperture decreases with increasing rock temperature, but the hydraulic conductivity increases owing to a reduction of the viscosity of the fluid flowing through. Finally, the heat-transfer efficiency of the samples under coupled hydro-thermal-mechanical conditions is analysed and discussed.

Experimental investigation of the hydraulic and heat-transfer properties of artificially fractured granite

PubMed Central

Luo, Jin; Zhu, Yongqiang; Guo, Qinghai; Tan, Long; Zhuang, Yaqin; Liu, Mingliang; Zhang, Canhai; Xiang, Wei; Rohn, Joachim

2017-01-01

In this paper, the hydraulic and heat-transfer properties of two sets of artificially fractured granite samples are investigated. First, the morphological information is determined using 3D modelling technology. The area ratio is used to describe the roughness of the fracture surface. Second, the hydraulic properties of fractured granite are tested by exposing samples to different confining pressures and temperatures. The results show that the hydraulic properties of the fractures are affected mainly by the area ratio, with a larger area ratio producing a larger fracture aperture and higher hydraulic conductivity. Both the hydraulic apertureand the hydraulic conductivity decrease with an increase in the confining pressure. Furthermore, the fracture aperture decreases with increasing rock temperature, but the hydraulic conductivity increases owing to a reduction of the viscosity of the fluid flowing through. Finally, the heat-transfer efficiency of the samples under coupled hydro-thermal-mechanical conditions is analysed and discussed. PMID:28054594

Multiphysics Code Demonstrated for Propulsion Applications

NASA Technical Reports Server (NTRS)

Lawrence, Charles; Melis, Matthew E.

1998-01-01

The utility of multidisciplinary analysis tools for aeropropulsion applications is being investigated at the NASA Lewis Research Center. The goal of this project is to apply Spectrum, a multiphysics code developed by Centric Engineering Systems, Inc., to simulate multidisciplinary effects in turbomachinery components. Many engineering problems today involve detailed computer analyses to predict the thermal, aerodynamic, and structural response of a mechanical system as it undergoes service loading. Analysis of aerospace structures generally requires attention in all three disciplinary areas to adequately predict component service behavior, and in many cases, the results from one discipline substantially affect the outcome of the other two. There are numerous computer codes currently available in the engineering community to perform such analyses in each of these disciplines. Many of these codes are developed and used in-house by a given organization, and many are commercially available. However, few, if any, of these codes are designed specifically for multidisciplinary analyses. The Spectrum code has been developed for performing fully coupled fluid, thermal, and structural analyses on a mechanical system with a single simulation that accounts for all simultaneous interactions, thus eliminating the requirement for running a large number of sequential, separate, disciplinary analyses. The Spectrum code has a true multiphysics analysis capability, which improves analysis efficiency as well as accuracy. Centric Engineering, Inc., working with a team of Lewis and AlliedSignal Engines engineers, has been evaluating Spectrum for a variety of propulsion applications including disk quenching, drum cavity flow, aeromechanical simulations, and a centrifugal compressor flow simulation.

Design and analysis of an active optics system for a 4-m telescope mirror combining hydraulic and pneumatic supports

NASA Astrophysics Data System (ADS)

Lousberg, Gregory P.; Moreau, Vincent; Schumacher, Jean-Marc; Piérard, Maxime; Somja, Aude; Gloesener, Pierre; Flebus, Carlo

2015-09-01

AMOS has developed a hybrid active optics system that combines hydraulic and pneumatic properties of actuators to support a 4-m primary mirror. The mirror is intended to be used in the Daniel K. Inouye Solar Telescope (DKIST, formerly the Advanced Technology Solar Telescope) that will be installed by the National Solar Observatory (NSO) atop the Haleakala volcano in Hawaii. The mirror support design is driven by the needs of (1) minimizing the support-induced mirror distortions under telescope operating conditions, (2) shaping the mirror surface to the desired profile, and (3) providing a high stiffness against wind loads. In order to fulfill these requirements, AMOS proposes an innovative support design that consist of 118 axial actuators and 24 lateral actuators. The axial support is based on coupled hydraulic and pneumatic actuators. The hydraulic part is a passive system whose main function is to support the mirror weight with a high stiffness. The pneumatic part is actively controlled so as to compensate for low-order wavefront aberrations that are generated by the mirror support itself or by any other elements in the telescope optical chain. The performances of the support and its adequacy with the requirements are assessed with the help of a comprehensive analysis loop involving finite-element, thermal and optical modellings.

Posttest analysis of the FFTF inherent safety tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Padilla, A. Jr.; Claybrook, S.W.

Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactormore » and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core and plant analysis was not completed in time to perform pretest analyses. The purpose of this paper is to present the results of the posttest analysis of the 1986 FFTF inherent safety tests using the SASSYS code.« less

Finite Element Modeling of Non-linear Coupled Interacting Fault System

NASA Astrophysics Data System (ADS)

Xing, H. L.; Zhang, J.; Wyborn, D.

2009-04-01

PANDAS - Parallel Adaptive static/dynamic Nonlinear Deformation Analysis System - a novel supercomputer simulation tool is developed for simulating the highly non-linear coupled geomechanical-fluid flow-thermal systems involving heterogeneously fractured geomaterials. PANDAS includes the following key components: Pandas/Pre, ESyS_Crustal, Pandas/Thermo, Pandas/Fluid and Pandas/Post as detailed in the following: • Pandas/Pre is developed to visualise the microseismicity events recorded during the hydraulic stimulation process to further evaluate the fracture location and evolution and geological setting of a certain reservoir, and then generate the mesh by it and/or other commercial graphics software (such as Patran) for the further finite element analysis of various cases; The Delaunay algorithm is applied as a suitable method for mesh generation using such a point set; • ESyS_Crustal is a finite element code developed for the interacting fault system simulation, which employs the adaptive static/dynamic algorithm to simulate the dynamics and evolution of interacting fault systems and processes that are relevant on short to mediate time scales in which several dynamic phenomena related with stick-slip instability along the faults need to be taken into account, i.e. (a). slow quasi-static stress accumulation, (b) rapid dynamic rupture, (c) wave propagation and (d) corresponding stress redistribution due to the energy release along the multiple fault boundaries; those are needed to better describe ruputure/microseimicity/earthquake related phenomena with applications in earthquake forecasting, hazard quantification, exploration, and environmental problems. It has been verified with various available experimental results[1-3]; • Pandas/Thermo is a finite element method based module for the thermal analysis of the fractured porous media; the temperature distribution is calculated from the heat transfer induced by the thermal boundary conditions without/with the coupled fluid effects and the geomechanical energy conversion for the pure/coupled thermal analysis. • Pandas/Fluid is a finite element method based module for simulating the fluid flow in the fractured porous media; the fluid flow velocity and pressure are calculated from energy equilibrium equations without/together with the coupling effects of the thermal and solid rock deformation for an independent/coupled fluid flow analysis; • Pandas/Post is to visualise the simulation results through the integration of VTK and/or Patran. All the above modules can be used independently/together to simulate individual/coupled phenomena (such as interacting fault system dynamics, heat flow and fluid flow) without/with coupling effects. PANDAS has been applied to the following issues: • visualisation of the microseismic events to monitor and determine where/how the underground rupture proceeds during a hydraulic stimulation, to generate the mesh using the recorded data for determining the domain of the ruptured zone and to evaluate the material parameters (i.e. the permeability) for the further numerical analysis; • interacting fault system simulation to determine the relevant complicated dynamic rupture process. • geomechanical-fluid flow coupling analysis to investigate the interactions between fluid flow and deformation in the fractured porous media under different loading conditions. • thermo-fluid flow coupling analysis of a fractured geothermal reservoir system. PANDAS will be further developed for a multiscale simulation of multiphase dynamic behaviour for a certain fractured geothermal reservoir. More details and additional application examples will be given during the presentation. References [1] Xing, H. L., Makinouchi, A. and Mora, P. (2007). Finite element modeling of interacting fault system, Physics of the Earth and Planetary Interiors, 163, 106-121.doi:10.1016/j.pepi.2007.05.006 [2] Xing, H. L., Mora, P., Makinouchi, A. (2006). An unified friction description and its application to simulation of frictional instability using finite element method. Philosophy Magazine, 86, 3453-3475 [3] Xing, H. L., Mora, P.(2006). Construction of an intraplate fault system model of South Australia, and simulation tool for the iSERVO institute seed project.. Pure and Applied Geophysics. 163, 2297-2316. DOI 10.1007/s00024-006-0127-x

Numerical modeling of consolidation processes in hydraulically deposited soils

NASA Astrophysics Data System (ADS)

Brink, Nicholas Robert

Hydraulically deposited soils are encountered in many common engineering applications including mine tailing and geotextile tube fills, though the consolidation process for such soils is highly nonlinear and requires the use of advanced numerical techniques to provide accurate predictions. Several commercially available finite element codes poses the ability to model soil consolidation, and it was the goal of this research to assess the ability of two of these codes, ABAQUS and PLAXIS, to model the large-strain, two-dimensional consolidation processes which occur in hydraulically deposited soils. A series of one- and two-dimensionally drained rectangular models were first created to assess the limitations of ABAQUS and PLAXIS when modeling consolidation of highly compressible soils. Then, geotextile tube and TSF models were created to represent actual scenarios which might be encountered in engineering practice. Several limitations were discovered, including the existence of a minimum preconsolidation stress below which numerical solutions become unstable.

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

A linearized microstructural model for hydraulic conductivity evolution due to brittle damage: application to Hydraulic Fracturing treatments

NASA Astrophysics Data System (ADS)

Caramiello, G.; Montanino, A.; Della Vecchia, G., Sr.; Pandolfi, A., Sr.

2017-12-01

Among the features of geological structures, fractures and discontinuities play a dominant role, due to their significant influence on both the hydraulic and the mechanical behavior of the rock mass. Despite the current availability of fault and fracture mappings, the understanding of the influence of faults on fluid flow is nowadays not satisfactory, in particular when hydro-mechanical coupling is significant. In engineering technology fracture processes are often exploited. Hydraulic fracturing is one of the most important example. Hydraulic fracturing is a process characterized by the inception and propagation of fractures as a consequence of a hydraulic driven solicitation and it is used to improve the production and optimize well stimulation in low permeability reservoirs. Due to the coupling of several different phenomena (hydro-thermo-chemical coupling) there is not a reliable complete mathematical model able to simulate in a proper way the process. To design hydraulic fracturing treatments, it is necessary to predict the growth of fracture geometry as a function of treatment parameters. In this contribution we present a recently developed model of brittle damage of confined rock masses, with particular emphasis on the influence of mechanical damage on the evolution of porosity and permeability. The model is based on an explicit micromechanical construction of connected patterns of parallel equi-spaced cracks. A relevant feature of the model is that the fracture patterns are not arbitrary, but their inception, orientation and spacing follow from energetic consideration. The model, based on the Terzaghi effective stress concepts, has been then implemented into a coupled hydro-mechanical finite element code, where the linear momentum and the fluid mass balance equations are numerically solved via a staggered approach. The coupled code is used to simulate fracturing processes induced by an increase in pore pressure. The examples show the capability of the model in reproducing three-dimensional multiscale complex fracture patterns and permeability enhancement in the damaged porous medium. The numerical code, has been used to verify the influence of the distance between the different perforation slots as well of the wellbore-deviation from the minimum stress axis on the propagation of multiple.

A fault constitutive relation accounting for thermal pressurization of pore fluid

USGS Publications Warehouse

Andrews, D.J.

2002-01-01

The heat generated in a slip zone during an earthquake can raise fluid pressure and thereby reduce frictional resistance to slip. The amount of fluid pressure rise depends on the associated fluid flow. The heat generated at a given time produces fluid pressure that decreases inversely with the square root of hydraulic diffusivity times the elapsed time. If the slip velocity function is crack-like, there is a prompt fluid pressure rise at the onset of slip, followed by a slower increase. The stress drop associated with the prompt fluid pressure rise increases with rupture propagation distance. The threshold propagation distance at which thermally induced stress drop starts to dominate over frictionally induced stress drop is proportional to hydraulic diffusivity. If hydraulic diffusivity is 0.02 m2/s, estimated from borehole samples of fault zone material, the threshold propagation distance is 300 m. The stress wave in an earthquake will induce an unknown amount of dilatancy and will increase hydraulic diffusivity, both of which will lessen the fluid pressure effect. Nevertheless, if hydraulic diffusivity is no more than two orders of magnitude larger than the laboratory value, then stress drop is complete in large earthquakes.

Analysis of the ITER central solenoid insert (CSI) coil stability tests

NASA Astrophysics Data System (ADS)

Savoldi, L.; Bonifetto, R.; Breschi, M.; Isono, T.; Martovetsky, N.; Ozeki, H.; Zanino, R.

2017-07-01

At the end of the test campaign of the ITER Central Solenoid Insert (CSI) coil in 2015, after 16,000 electromagnetic (EM) cycles, some tests were devoted to the study of the conductor stability, through the measurement of the Minimum Quench Energy (MQE). The tests were performed by means of an inductive heater (IH), located in the high-field region of the CSI and wrapped around the conductor. The calorimetric calibration of the IH is presented here, aimed at assessing the energy deposited in the conductor for different values of the IH electrical operating conditions. The MQE of the conductor of the ITER CS module 3L can be estimated as ∼200 J ± 20%, deposited on the whole conductor on a length of ∼10 cm (the IH length) in ∼40 ms, at current and magnetic field conditions relevant for the ITER CS operation. The repartition of the energy deposited in the conductor under the IH is computed to be ∼10% in the cable and 90% in the jacket by means of a 3D Finite Elements EM model. It is shown how this repartition implies that the bundle (cable + helium) heat capacity is fully available for stability on the time scale of the tested disturbances. This repartition is used in input to the thermal-hydraulic analysis performed with the 4C code, to assess the capability of the model to accurately reproduce the stability threshold of the conductor. The MQE computed by the code for this disturbance is in good agreement with the measured value, with an underestimation within 15% of the experimental value.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

User's manual for the one-dimensional hypersonic experimental aero-thermodynamic (1DHEAT) data reduction code

NASA Technical Reports Server (NTRS)

Hollis, Brian R.

1995-01-01

A FORTRAN computer code for the reduction and analysis of experimental heat transfer data has been developed. This code can be utilized to determine heat transfer rates from surface temperature measurements made using either thin-film resistance gages or coaxial surface thermocouples. Both an analytical and a numerical finite-volume heat transfer model are implemented in this code. The analytical solution is based on a one-dimensional, semi-infinite wall thickness model with the approximation of constant substrate thermal properties, which is empirically corrected for the effects of variable thermal properties. The finite-volume solution is based on a one-dimensional, implicit discretization. The finite-volume model directly incorporates the effects of variable substrate thermal properties and does not require the semi-finite wall thickness approximation used in the analytical model. This model also includes the option of a multiple-layer substrate. Fast, accurate results can be obtained using either method. This code has been used to reduce several sets of aerodynamic heating data, of which samples are included in this report.

Numerical predictions of EML (electromagnetic launcher) system performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schnurr, N.M.; Kerrisk, J.F.; Davidson, R.F.

1987-01-01

The performance of an electromagnetic launcher (EML) depends on a large number of parameters, including the characteristics of the power supply, rail geometry, rail and insulator material properties, injection velocity, and projectile mass. EML system performance is frequently limited by structural or thermal effects in the launcher (railgun). A series of computer codes has been developed at the Los Alamos National Laboratory to predict EML system performance and to determine the structural and thermal constraints on barrel design. These codes include FLD, a two-dimensional electrostatic code used to calculate the high-frequency inductance gradient and surface current density distribution for themore » rails; TOPAZRG, a two-dimensional finite-element code that simultaneously analyzes thermal and electromagnetic diffusion in the rails; and LARGE, a code that predicts the performance of the entire EML system. Trhe NIKE2D code, developed at the Lawrence Livermore National Laboratory, is used to perform structural analyses of the rails. These codes have been instrumental in the design of the Lethality Test System (LTS) at Los Alamos, which has an ultimate goal of accelerating a 30-g projectile to a velocity of 15 km/s. The capabilities of the individual codes and the coupling of these codes to perform a comprehensive analysis is discussed in relation to the LTS design. Numerical predictions are compared with experimental data and presented for the LTS prototype tests.« less

Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

2006-06-01

Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Analysis of Anderson Acceleration on a Simplified Neutronics/Thermal Hydraulics System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toth, Alex; Kelley, C. T.; Slattery, Stuart R

ABSTRACT A standard method for solving coupled multiphysics problems in light water reactors is Picard iteration, which sequentially alternates between solving single physics applications. This solution approach is appealing due to simplicity of implementation and the ability to leverage existing software packages to accurately solve single physics applications. However, there are several drawbacks in the convergence behavior of this method; namely slow convergence and the necessity of heuristically chosen damping factors to achieve convergence in many cases. Anderson acceleration is a method that has been seen to be more robust and fast converging than Picard iteration for many problems, withoutmore » significantly higher cost per iteration or complexity of implementation, though its effectiveness in the context of multiphysics coupling is not well explored. In this work, we develop a one-dimensional model simulating the coupling between the neutron distribution and fuel and coolant properties in a single fuel pin. We show that this model generally captures the convergence issues noted in Picard iterations which couple high-fidelity physics codes. We then use this model to gauge potential improvements with regard to rate of convergence and robustness from utilizing Anderson acceleration as an alternative to Picard iteration.« less

Use of Generalized Fluid System Simulation Program (GFSSP) for Teaching and Performing Senior Design Projects at the Educational Institutions

NASA Technical Reports Server (NTRS)

Majumdar, A. K.; Hedayat, A.

2015-01-01

This paper describes the experience of the authors in using the Generalized Fluid System Simulation Program (GFSSP) in teaching Design of Thermal Systems class at University of Alabama in Huntsville. GFSSP is a finite volume based thermo-fluid system network analysis code, developed at NASA/Marshall Space Flight Center, and is extensively used in NASA, Department of Defense, and aerospace industries for propulsion system design, analysis, and performance evaluation. The educational version of GFSSP is freely available to all US higher education institutions. The main purpose of the paper is to illustrate the utilization of this user-friendly code for the thermal systems design and fluid engineering courses and to encourage the instructors to utilize the code for the class assignments as well as senior design projects.

TOPAZ2D heat transfer code users manual and thermal property data base

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shapiro, A.B.; Edwards, A.L.

1990-05-01

TOPAZ2D is a two dimensional implicit finite element computer code for heat transfer analysis. This user's manual provides information on the structure of a TOPAZ2D input file. Also included is a material thermal property data base. This manual is supplemented with The TOPAZ2D Theoretical Manual and the TOPAZ2D Verification Manual. TOPAZ2D has been implemented on the CRAY, SUN, and VAX computers. TOPAZ2D can be used to solve for the steady state or transient temperature field on two dimensional planar or axisymmetric geometries. Material properties may be temperature dependent and either isotropic or orthotropic. A variety of time and temperature dependentmore » boundary conditions can be specified including temperature, flux, convection, and radiation. Time or temperature dependent internal heat generation can be defined locally be element or globally by material. TOPAZ2D can solve problems of diffuse and specular band radiation in an enclosure coupled with conduction in material surrounding the enclosure. Additional features include thermally controlled reactive chemical mixtures, thermal contact resistance across an interface, bulk fluid flow, phase change, and energy balances. Thermal stresses can be calculated using the solid mechanics code NIKE2D which reads the temperature state data calculated by TOPAZ2D. A three dimensional version of the code, TOPAZ3D is available. The material thermal property data base, Chapter 4, included in this manual was originally published in 1969 by Art Edwards for use with his TRUMP finite difference heat transfer code. The format of the data has been altered to be compatible with TOPAZ2D. Bob Bailey is responsible for adding the high explosive thermal property data.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

NASA Astrophysics Data System (ADS)

Shadid, J. N.; Smith, T. M.; Cyr, E. C.; Wildey, T. M.; Pawlowski, R. P.

2016-09-01

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier-Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shadid, J.N., E-mail: jnshadi@sandia.gov; Department of Mathematics and Statistics, University of New Mexico; Smith, T.M.

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts tomore » apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shadid, J. N.; Smith, T. M.; Cyr, E. C.

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. The understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In our study we report on initial efforts to apply integrated adjoint-basedmore » computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. We present the initial results that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

DOE PAGES

Shadid, J. N.; Smith, T. M.; Cyr, E. C.; ...

2016-05-20

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. The understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In our study we report on initial efforts to apply integrated adjoint-basedmore » computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. We present the initial results that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.« less

On the collaborative design and simulation of space camera: stop structural/thermal/optical) analysis

NASA Astrophysics Data System (ADS)

Duan, Pengfei; Lei, Wenping

2017-11-01

A number of disciplines (mechanics, structures, thermal, and optics) are needed to design and build Space Camera. Separate design models are normally constructed by each discipline CAD/CAE tools. Design and analysis is conducted largely in parallel subject to requirements that have been levied on each discipline, and technical interaction between the different disciplines is limited and infrequent. As a result a unified view of the Space Camera design across discipline boundaries is not directly possible in the approach above, and generating one would require a large manual, and error-prone process. A collaborative environment that is built on abstract model and performance template allows engineering data and CAD/CAE results to be shared across above discipline boundaries within a common interface, so that it can help to attain speedy multivariate design and directly evaluate optical performance under environment loadings. A small interdisciplinary engineering team from Beijing Institute of Space Mechanics and Electricity has recently conducted a Structural/Thermal/Optical (STOP) analysis of a space camera with this collaborative environment. STOP analysis evaluates the changes in image quality that arise from the structural deformations when the thermal environment of the camera changes throughout its orbit. STOP analyses were conducted for four different test conditions applied during final thermal vacuum (TVAC) testing of the payload on the ground. The STOP Simulation Process begins with importing an integrated CAD model of the camera geometry into the collaborative environment, within which 1. Independent thermal and structural meshes are generated. 2. The thermal mesh and relevant engineering data for material properties and thermal boundary conditions are then used to compute temperature distributions at nodal points in both the thermal and structures mesh through Thermal Desktop, a COTS thermal design and analysis code. 3. Thermally induced structural deformations of the camera are then evaluated in Nastran, an industry standard code for structural design and analysis. 4. Thermal and structural results are next imported into SigFit, another COTS tool that computes deformation and best fit rigid body displacements for the optical surfaces. 5. SigFit creates a modified optical prescription that is imported into CODE V for evaluation of optical performance impacts. The integrated STOP analysis was validated using TVAC test data. For the four different TVAC tests, the relative errors between simulation and test data of measuring points temperatures were almost around 5%, while in some test conditions, they were even much lower to 1%. As to image quality MTF, relative error between simulation and test was 8.3% in the worst condition, others were all below 5%. Through the validation, it has been approved that the collaborative design and simulation environment can achieved the integrated STOP analysis of Space Camera efficiently. And further, the collaborative environment allows an interdisciplinary analysis that formerly might take several months to perform to be completed in two or three weeks, which is very adaptive to scheme demonstration of projects in earlier stages.

Rocketdyne/Westinghouse nuclear thermal rocket engine modeling

NASA Technical Reports Server (NTRS)

Glass, James F.

1993-01-01

The topics are presented in viewgraph form and include the following: systems approach needed for nuclear thermal rocket (NTR) design optimization; generic NTR engine power balance codes; rocketdyne nuclear thermal system code; software capabilities; steady state model; NTR engine optimizer code-logic; reactor power calculation logic; sample multi-component configuration; NTR design code output; generic NTR code at Rocketdyne; Rocketdyne NTR model; and nuclear thermal rocket modeling directions.

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pahari, S.; Hajela, S.; Rammohan, H. P.

2012-07-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG andmore » PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)« less

AN OVERVIEW OF REDUCED ORDER MODELING TECHNIQUES FOR SAFETY APPLICATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, D.; Alfonsi, A.; Talbot, P.

2016-10-01

The RISMC project is developing new advanced simulation-based tools to perform Computational Risk Analysis (CRA) for the existing fleet of U.S. nuclear power plants (NPPs). These tools numerically model not only the thermal-hydraulic behavior of the reactors primary and secondary systems, but also external event temporal evolution and component/system ageing. Thus, this is not only a multi-physics problem being addressed, but also a multi-scale problem (both spatial, µm-mm-m, and temporal, seconds-hours-years). As part of the RISMC CRA approach, a large amount of computationally-expensive simulation runs may be required. An important aspect is that even though computational power is growing, themore » overall computational cost of a RISMC analysis using brute-force methods may be not viable for certain cases. A solution that is being evaluated to assist the computational issue is the use of reduced order modeling techniques. During the FY2015, we investigated and applied reduced order modeling techniques to decrease the RISMC analysis computational cost by decreasing the number of simulation runs; for this analysis improvement we used surrogate models instead of the actual simulation codes. This article focuses on the use of reduced order modeling techniques that can be applied to RISMC analyses in order to generate, analyze, and visualize data. In particular, we focus on surrogate models that approximate the simulation results but in a much faster time (microseconds instead of hours/days).« less

Some selected quantitative methods of thermal image analysis in Matlab.

PubMed

Koprowski, Robert

2016-05-01

The paper presents a new algorithm based on some selected automatic quantitative methods for analysing thermal images. It shows the practical implementation of these image analysis methods in Matlab. It enables to perform fully automated and reproducible measurements of selected parameters in thermal images. The paper also shows two examples of the use of the proposed image analysis methods for the area of ​​the skin of a human foot and face. The full source code of the developed application is also provided as an attachment. The main window of the program during dynamic analysis of the foot thermal image. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Reliability of Next Generation Power Electronics Packaging Under Concurrent Vibration, Thermal and High Power Loads

DTIC Science & Technology

2008-02-01

combined thermal g effect and initial current field. The model is implemented using Abaqus user element subroutine and verified against the experimental...Finite Element Formulation The proposed model is implemented with ABAQUS general purpose finite element program using thermal -displacement analysis...option. ABAQUS and other commercially available finite element codes do not have the capability to solve general electromigration problem directly. Thermal

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

Pulsed thermography detection of water and hydraulic oil intrusion in the honeycomb sandwich structure composite

NASA Astrophysics Data System (ADS)

Zhao, Shi-bin; Zhang, Cun-lin; Wu, Nai-ming

2011-08-01

Water and hydraulic oil intrusion inside honeycomb sandwich Structure Composite during service has been linked to in-flight failure in some aircraft. There is an ongoing effort to develop nondestructive testing methods to detect the presence of water and hydraulic oil within the sandwich panels. Pulsed thermography(PT) represents an attractive approach in that it is sensitive to the change of thermal properties. Using a flash lamp PT, testing can be applied directly to the surface of the panel. The viability of PT is demonstrated through laboratory imaging of both water and hydraulic oil within sandwich panels. The detection of water and hydraulic oil intrusion using a one-sided flash lamp PT is presented. It is shown that simple detection, as well as spatial localization of water and hydraulic oil within sandwich panels, and assign the quantity of water and hydraulic oil is possible.

CFD Aided Design and Production of Hydraulic Turbines

NASA Astrophysics Data System (ADS)

Kaplan, Alper; Cetinturk, Huseyin; Demirel, Gizem; Ayli, Ece; Celebioglu, Kutay; Aradag, Selin; ETU Hydro Research Center Team

2014-11-01

Hydraulic turbines are turbo machines which produce electricity from hydraulic energy. Francis type turbines are the most common one in use today. The design of these turbines requires high engineering effort since each turbine is tailor made due to different head and discharge. Therefore each component of the turbine is designed specifically. During the last decades, Computational Fluid Dynamics (CFD) has become very useful tool to predict hydraulic machinery performance and save time and money for designers. This paper describes a design methodology to optimize a Francis turbine by integrating theoretical and experimental fundamentals of hydraulic machines and commercial CFD codes. Specific turbines are designed and manufactured with the help of a collaborative CFD/CAD/CAM methodology based on computational fluid dynamics and five-axis machining for hydraulic electric power plants. The details are presented in this study. This study is financially supported by Turkish Ministry of Development.

Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Seung Jun; Buechler, Cynthia Eileen

The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operatingmore » scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi-physics methodology and preliminary results from various coupled calculations (power prediction and heat transfer coefficient) can be further utilized for the system code validation and generic solution vessel design improvement.« less