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Title: Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code

Abstract

A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature. New and safer nuclear reactors (Generation IV) are now in the early planning stages in many countries throughout the world. One of the reactor concepts being seriously considered is the VHTR. To achieve public acceptance, these reactor concepts must show an increased level of inherent safety over current reactor designs (i.e., a system must be designed to eliminate any concerns of large radiological releases outside the site boundary). A computer code developed from this study, gas multi-component mixture analysis (GAMMA) code, was assessed using a two-bulb experiment and in addition the molecular diffusion behavior in the prismatic-core gas-cooled reactor was investigated following the guillotine break of the main pipe between the reactor vessel andmore » the power conversion unit. The RELAP5 code was improved for the VHTR air ingress analysis and was assessed using inverse U-tube and NACOK natural circulation data.« less

Authors:
Publication Date:
Research Org.:
Idaho National Laboratory (INL)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
915516
Report Number(s):
INL/JOU-06-11845
TRN: US0804998
DOE Contract Number:
DE-AC07-99ID-13727
Resource Type:
Journal Article
Resource Relation:
Journal Name: Nuclear Engineering & Design; Journal Volume: 237; Journal Issue: 10
Country of Publication:
United States
Language:
English
Subject:
99 - GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE; AIR; CHEMICAL REACTIONS; COMPUTER CODES; DIFFUSION; GAS COOLED REACTORS; LOSS OF COOLANT; NATURAL CONVECTION; REACTOR VESSELS; VHTR, thermal hydraulics

Citation Formats

Chang Oh. Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code. United States: N. p., 2007. Web.
Chang Oh. Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code. United States.
Chang Oh. Tue . "Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code". United States. doi:.
@article{osti_915516,
title = {Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code},
author = {Chang Oh},
abstractNote = {A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature. New and safer nuclear reactors (Generation IV) are now in the early planning stages in many countries throughout the world. One of the reactor concepts being seriously considered is the VHTR. To achieve public acceptance, these reactor concepts must show an increased level of inherent safety over current reactor designs (i.e., a system must be designed to eliminate any concerns of large radiological releases outside the site boundary). A computer code developed from this study, gas multi-component mixture analysis (GAMMA) code, was assessed using a two-bulb experiment and in addition the molecular diffusion behavior in the prismatic-core gas-cooled reactor was investigated following the guillotine break of the main pipe between the reactor vessel and the power conversion unit. The RELAP5 code was improved for the VHTR air ingress analysis and was assessed using inverse U-tube and NACOK natural circulation data.},
doi = {},
journal = {Nuclear Engineering & Design},
number = 10,
volume = 237,
place = {United States},
year = {Tue May 01 00:00:00 EDT 2007},
month = {Tue May 01 00:00:00 EDT 2007}
}
  • An experience with the vectorization of the light water reactor transient analysis code RELAP5/MOD1 on a vector supercomputer FACOM VP-100 (peak speed 250 million floating point operations/s, clock period 7.5 ns) is described. The approach to the vectorization is based on the junction and volume level parallelisms for the hydrodynamic model, and the heat structure and heat mesh levels for the heat transfer model. The VP-100 vectorized code version yields a 2.4 to 2.8 factor speed increase over the FACOM M-380 computer, depending on the number of spatial cells being used. The M-380 is an IBM-type computer with the samemore » speed as the VP-100 in scalar mode.« less
  • A computational investigation of experiments involving the condensation phenomenon in the presence of noncondensable gases was performed. The RELAP5/MOD3 thermal-hydraulic code was utilized for this analysis. Two separate-effects experiments were studied, which are relevant to actual situations encountered in the industry. The first experiment involved condensation of steam in an inverted U-tube when nitrogen is present. A constant flow of steam was injected into the U-tube and condensed along its surface. The condensing length was a function of the injected nitrogen rate and the secondary temperature. The code predicted an active condensation zone with unimpeded heat transfer and a passivemore » zone with no heat transfer. The lengths of these zones agree with the experimental data. The gas temperatures in the U-tube were favorably predicted except for a discrepancy where the calculated primary temperatures were lower than the secondary temperatures for several cases. Active nitrogen contents in the tube were underpredicted by the code. The second experiment investigated was the Massachusetts Institute of Technology's steam condensation experiment. This experiment modeled the proposed containment cooling system for advanced reactors. Steam was generated in a vessel in which air was present. The steam in the steam-air mixture condensed on the surface of a cooled copper cylinder. Computational predictions of this experiment revealed that heat transfer coefficients vary with air fraction. Calculated heat transfer coefficients were compared with the data, and it was found that the results were better for higher system pressures than for lower pressures.« less
  • This work presents the implementation of the Adjoint Sensitivity Analysis Procedure for the nonequilibrium, nonhomogeneous two-fluid model, including boron concentration and noncondensable gases, of the RELAP5/MOD3.2 code. The end product of this implementation is the Adjoint Sensitivity Model (ASM-REL/TF), which is derived for both the differential and discretized equations underlying the two-fluid model. The consistency requirements between these two representations are also highlighted.
  • This work presents results that illustrate the validation of the Adjoint Sensitivity Model (ASM-REL/TF) corresponding to the two-fluid model with noncondensable(s) used in RELAP5/MOD3.2. This validation has been carried out by using sample problems involving (a) a liquid phase only, (b) a gas phase only, and (c) a two-phase mixture (of water and steam). Thus, the 'Two-Loops with Pumps' sample problem supplied with RELAP5/MOD3.2 has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when only the liquid phase is present. Furthermore, the 'Edwards Pipe' sample problem, also supplied with RELAP5/MOD3.2, has been usedmore » to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when both (i.e., liquid and gas) phases are present. In addition, the accuracy and stability have been verified of the numerical solution of the ASM-REL/TF when only the gas phase is present by using modified 'Two-Loops with Pumps' and the 'Edwards Pipe' sample problems in which the liquid- and two-phase fluids, respectively, were replaced with pure steam. The results obtained for these sample problems depict typical sensitivities of junction velocities and volume-averaged pressures to perturbations in initial conditions and indicate that the numerical solution of the ASM-REL/TF is as robust, stable, and accurate as the original RELAP5/MOD3.2 calculations.This work also illustrates the role that sensitivities of the thermodynamic properties of water play for sensitivity analysis of thermal-hydraulic codes for light water reactors. The well-known 1993 ASME Steam Tables are used to present typical analytical and numerical results for sensitivities of the thermodynamic properties of water to the numerical parameters that appear in the mathematical formulation of these properties. Particularly highlighted are the very large sensitivities displayed by the specific isobaric fluid and gas heat capacities C{sub pf} and C{sub pg}, respectively; the specific fluid enthalpy h{sub f}; the specific gas volume V{sub g}; the volumetric expansion coefficient for gas {beta}{sub g}; and the isothermal coefficient for gas k{sub g}. The dependence of {beta}{sub g} and k{sub g} on the most sensitive parameters turns out to be nonlinear, while the dependence of C{sub pf}, C{sub pg}, h{sub f}, and V{sub g} on the most sensitive parameters turns out to be linear, so the respective sensitivities predict exactly the effects of variations in the respective parameters. On the other hand, the sensitivities of the specific fluid volume V{sub f}, the volumetric expansion coefficient for fluid {beta}{sub f}, the specific gas enthalpy h{sub g}, and the isothermal coefficient of compressibility for fluid k{sub f} to the parameters that appear in their respective mathematical formulas are quite small. Finally, it is noted that such deterministically calculated sensitivities can be used to rank the respective parameters according to their importance, to assess the effects of nonlinearities and, more generally, to perform comprehensive sensitivity/uncertainty analyses of thermal-hydraulic codes that use a water substance as the working fluid.« less
  • The adjoint sensitivity analysis procedure for augmented systems for application to the RELAP5/MOD3.2 code system is illustrated. Specifically, the adjoint sensitivity model corresponding to the heat structure models in RELAP5/MOD3.2 is derived and subsequently augmented to the two-fluid adjoint sensitivity model (ASM-REL/TF). The end product, called ASM-REL/TFH, comprises the complete adjoint sensitivity model for the coupled fluid dynamics/heat structure packages of the large-scale simulation code RELAP5/MOD3.2. The ASM-REL/TFH model is validated by computing sensitivities to the initial conditions for various time-dependent temperatures in the test bundle of the Quench-04 reactor safety experiment. This experiment simulates the reflooding with water ofmore » uncovered, degraded fuel rods, clad with material (Zircaloy-4) that has the same composition and size as that used in typical pressurized water reactors. The most important response for the Quench-04 experiment is the time evolution of the cladding temperature of heated fuel rods. The ASM-REL/TFH model is subsequently used to perform an illustrative sensitivity analysis of this and other time-dependent temperatures within the bundle. The results computed by using the augmented adjoint sensitivity system, ASM-REL/TFH, highlight the reliability, efficiency, and usefulness of the adjoint sensitivity analysis procedure for computing time-dependent sensitivities.« less