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Implementation of Molten Salt Properties into RELAP5-3D/ATHENA

Molten salts are being considered as coolants for the Next Generation Nuclear Plant (NGNP) in both the reactor and the heat transport loop between the reactor and the hydrogen production plant because of their superior thermophysical properties compared to helium. Because specific molten salts have not been selected for either application, four separate molten salts were implemented into the RELAP5-3D/ATHENA computer program as working fluids. The implemented salts were LiF-BeF2 in a molar mixture that is 66% LiF and 34% BeF2, respectively, NaBF4-NaF (92% and 8%), LiF-NaF-KF (11.5%, 46.5%, and 42%), and NaF-ZrF4 (50% and 50%). LiF-BeF2 is currently the first choice for the primary coolant for the Advanced High- Temperature Reactor, while NaF-ZrF4 is being considered as an alternate. NaBF4-NaF and LiFNaF- KF are being considered as possible coolants for the heat transport loop. The molten salts were implemented into ATHENA using a simplified equation of state based on data and correlations obtained from Oak Ridge National Laboratory. The simplified equation of state assumes that the liquid density is a function of temperature and pressure and that the liquid heat capacity is constant. The vapor is assumed to have the same composition as the liquid and is assumed to be a perfect gas. The implementation of the thermodynamic properties into ATHENA for LiF-BeF2 was verified by comparisons with results from a detailed equation of state that utilized a soft-sphere model. The comparisons between the simplified and soft-sphere models were in reasonable agreement for liquid. The agreement for vapor properties was not nearly as good as that obtained for liquid. Large uncertainties are possible in the vapor properties because of a lack of experimental data. The simplified model used here is not expected to be accurate for boiling or single-phase vapor conditions. Because neither condition is expected during NGNP applications, the simplified equation of state is considered acceptably accurate for analysis of the NGNP

Building an Institutional Repository in Hard Times

This poster presents an overview of an exploratory research initiative to examine and assess the viability of developing an institutional repository system at a teaching-oriented four-year university with minimal monetary commitment. A need has been identified for an institutional repository and necessary steps have been taken to implement it. Several departments worked together to create a prototype Institutional Repository using DSpace, an open source repository software. This repository represents a unique endeavor, in that it has been instituted at a non-research based university and has chosen to involve students in the planning, design, implementation, and documentation stages of the project. In addition, the university\u27s Library Science students will also be involved in creating and maintaining collections. This poster focuses on the steps taken to set up and the plans to maintain a quality Institutional Repository at Valdosta State University without placing a large demand on the institution\u27s resources

Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor

Analysis Methods and Desired Outcomes of System Interface Heat Transfer Fluid Requirements and Characteristics Analyses

The interface between the Next Generation Nuclear Plant (NGNP) and the hydrogen-generating process plant will contain an intermediate loop that will transport heat from the NGNP to the process plant. Seven possible configurations for the NGNP primary coolant system and the intermediate heat transport loop were identified. Both helium and liquid salts are being considered as the working fluid in the intermediate heat transport loop. A method was developed to perform thermal-hydraulic evaluations of the different configurations and coolants. The evaluations will determine which configurations and coolants are the most promising from a thermal-hydraulic point of view and which, if any, do not appear to be feasible at the current time. Results of the evaluations will be presented in a subsequent report

Verification and Validation of Corrected Versions of RELAP5 for ATR Reactivity Analyses

Two versions of the RELAP5 computer code, RELAP5/MOD2.5 and RELAP5/MOD3 Version 3.2.1.2, are used to support safety analyses of the Advanced Test Reactor (ATR). Both versions of RELAP5 contain a point reactor kinetics model that has been used to simulate power excursion transients at the ATR. Errors in the RELAP5 point kinetics model were reported to the RELAP5 code developers in 2007. These errors had the potential to affect reactivity analyses that are part of the ATR’s safety basis. Consequently, corrected versions of RELAP5 were developed for analysis of the ATR. Four reactivity transients were simulated to verify and validate the corrected codes for use in safety evaluations of the ATR. The objectives of this paper are to describe the verification and validation of the point kinetics model for ATR applications and to inform code users of the effects of the errors on representative reactivity analyses

RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment

Thermal Response of the Hybrid Loop-Pool Design for Sodium Cooled Faster Reactors

An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unproctected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems

Thermal-Hydraulic Analyses Of The LS-VHTR

Thermal-hydraulic analyses were performed to evaluate the safety characteristics of the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). A one-dimensional model of the LS-VHTR was developed using the RELAP5-3D computer program. The thermal calculations from the one-dimensional model of a fuel block were benchmarked against a multi-dimensional finite element model. The RELAP5-3D model was used to simulate a transient initiated by loss of forced convection in which the Reactor Vessel Auxiliary Cooling System (RVACS) passively removed decay heat. Parametric calculations were performed to investigate the effects of various parameters, including bypass flow fraction, coolant channel diameter, and the coolant outlet temperature. Additional parametric calculations investigated the effects of an enhanced RVACS design, failure to scram, and radial/axial conduction in the core

Assessment of RELAP5-3D for Analysis of Very High Temperature Gas-Cooled Reactors

The RELAP5-3D© computer code is being improved for the analysis of very high temperature gas-cooled reactors. Diffusion and natural circulation can be important phenomena in gas-cooled reactors following a loss-of-coolant accident. Recent improvements to the code include the addition of models that simulate pressure loss across a pebble bed and molecular diffusion. These models were assessed using experimental data. The diffusion model was assessed using data from inverted U-tube experiments. The code’s capability to simulate natural circulation of air through a pebble bed was assessed using data from the NACOK facility. The calculated results were in reasonable agreement with the measured values