Implementation of Liquid Metal Properties in RELAP5 MOD3.2 for Safety Analysis of Sodium-Cooled Fast Reactors

2017 ◽  
Author(s):  
Jian Song ◽  
Limin Liu ◽  
Simiao Tang ◽  
Yingwei Wu ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Experimental Data ◽  
Liquid Metal ◽  
Fast Reactor ◽  
Natural Circulation ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Working Fluid ◽  
Primary System ◽  
Steady State Condition ◽  
Design Values

Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the modified RELAP5 MOD3.2 is being considered as the thermal-hydraulic system code to support the development of the SFRs. The thermodynamic and transport properties of sodium liquid and vapor have been implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The sodium liquid properties use polynomial equations based on data obtained from Argonne National Laboratory, and the vapor is assumed to be perfect gas. The property equations are acceptably accurate for analysis of SFR, especially for single-phase liquid. New files are added to the fluids directory to generate property tables for new working fluid, which are similar to the table interpolation subroutines for light and heavy water in the original file directory. The method of code modifications are universal for other working fluids and will not affect the code original performance. Some basic verification work for the modified code are carried out. The steam generator of CEFR is analyzed to verify the modified code. The calculated results show that all the water will boil off in the evaporator and the calculated results are in good agreement with the design values. By using modified RELAP5 to model the primary loop of EBR-II fast reactor, the SHRT-17 PLOF test was analyzed. The results show that the natural circulation can be established in the EBR-II primary system after main pumps off to remove the core decay residual heat effectively, and the peak temperature under the safety limits. Moreover, the results computed in this work compared well with the test experimental data for the steady state condition. During the transients, the changing trends of temperature and pressure are similar to experimental data. The discrepancies between calculation and experiment are considered acceptably which need to be improved in the future work. Our work could demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.

scholarly journals Shutdown and Closure of the Experimental Breeder Reactor–II

2002 ◽  
Author(s):  
John A. Michelbacher ◽  
Carl E. Baily ◽  
Daniel K. Baird ◽  
S. Paul Henslee ◽  
Collin J. Knight ◽  
...  
Keyword(s):  
Liquid Metal ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Passive Layer ◽  
Department Of Energy ◽  
Primary System ◽  
Secondary System ◽  
Reactor Plant ◽  
Carbon Dioxide Gas ◽  
Safe Condition

The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.

Performance of Sodium Natural Circulation Loop for Passvie Decay Heat Removal

2008 ◽  
Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  
Keyword(s):  
Fast Reactor ◽  
System Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Circulation Loop ◽  
Ambient Atmosphere ◽  
Steady State Condition ◽  
Decay Heat ◽  
Natural Circulation Loop ◽  
Test Loop

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.

Heat Exchanger Shellside Pressure Drop: Comparison of Predictions With Experimental Data

1988 ◽  
Vol 110 (1) ◽  
pp. 68-76 ◽  
Author(s):  
R. S. Kistler ◽  
J. M. Chenoweth
Keyword(s):  
Experimental Data ◽  
Heat Exchanger ◽  
Pressure Drop ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Flow Induced Vibration ◽  
Ongoing Research ◽  
Pressure Drops ◽  
Complementary Method ◽  
Nozzle Pressure

A unique set of heat exchanger shellside pressure drop experimental data has become available from experiments at Argonne National Laboratory as a part of an ongoing research program in flow-induced vibration. These data provide overall pressure drop for a number of typical industrial heat exchanger configurations in addition to incremental pressure drop measurements along the shellside path. The test program systematically varied the baffle spacing, the tubefield pattern, and nozzle size for a series of isothermal water tests for segmentally baffled bundles. Also recently a comprehensive method has been published in the Heat Exchanger Design Handbook (HEDH) for the prediction of bundle shellside pressure drops. A search of the literature failed to reveal a complementary method for predicting the shellside nozzle pressure losses. This paper compares the predicted with the measured data and validates the adequacy and limitations of the HEDH method for full bundles of plain tubes. It further applies an extension to the method for no-tubes-in-the-window bundles. Adjustments were indicated to improve the predictions for finned tubes and methods were developed to predict shellside nozzle pressure drops. Overall pressure drop predictions were within plus or minus 20 percent.

scholarly journals A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

2016 ◽  
Author(s):  
Dave Grabaskas ◽  
Acacia J. Brunett ◽  
Matthew Bucknor
Keyword(s):  
Source Term ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Primary System ◽  
Separate Analysis ◽  
Event Sequences ◽  
Safety And Reliability ◽  
Detailed Assessment ◽  
Water Reactors ◽  
High Level

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release of radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.

Supercritical Natural Circulation Loop

2021 ◽  
pp. 338-369
Author(s):  
Tanuj Srivastava ◽  
Pranab Sutradhar ◽  
Milan Krishna Singha Sarkar ◽  
Dipankar Narayan Basu
Keyword(s):  
Nuclear Reactors ◽  
Natural Circulation ◽  
Point Of View ◽  
Circulation Loop ◽  
Working Fluid ◽  
System Behavior ◽  
Steady State Condition ◽  
Natural Circulation Loop ◽  
Stability Point ◽  
Future Direction

Supercritical natural circulation loop is a compelling technology for cooling of modern nuclear reactors, which promises enhanced thermal-hydraulic performance in a simple design. Being a new concept, related knowledge base is relatively thin and involves several conflicting theories and controversies. The chapter summarizes the observation till date, starting from the very fundamentals. The phenomenon of natural circulation under steady state condition and suitability of supercritical medium as working fluid are discussed in detail. Different methods of analyses, including analytical, simple 1-d numerical, and multidimensional computational codes, as well as experimental, are elucidated. A comprehensive discussion is presented about the effect of various geometric and operating parameters on the system behavior, from both thermal-hydraulic and stability point of view. Finally, a few recommendations are included about the operation of such loops and future direction of research.

Comparative Study on Neutronics Characteristics of a 1500 MWe Metal Fuel Sodium-Cooled Fast Reactor

2016 ◽  
Author(s):  
Kazuya Ohgama ◽  
Gerardo Aliberti ◽  
Nicolas E. Stauff ◽  
Shigeo Ohki ◽  
Taek K. Kim
Keyword(s):  
Fast Reactor ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Nuclear Data ◽  
Cooperative Effort ◽  
Energy Agency ◽  
Control Rod ◽  
Benchmark Study ◽  
Metal Fuel ◽  
And Control

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions are agreed well with less than 200 pcm of discrepancy on the neutron multiplication factor, and less than 3% of discrepancy on the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology.

Evaluation of Advanced Thermal-Hydraulic System Codes for Safety Analysis of Integral Type PWR

2013 ◽  
Author(s):  
J. Choi ◽  
B. Woods
Keyword(s):  
Experimental Data ◽  
Boundary Conditions ◽  
Hydraulic System ◽  
System Analysis ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Test Facility ◽  
Primary System ◽  
Pressurized Water ◽  
Water Reactor

The integral Pressurized Water Reactor (PWR) concept, which contains the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high possibility for near-term deployment. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Primary System and Containment during Accidents” has been conducted since 2010. Oregon State University of USA has offered their experimental facility, which was built to demonstrate the feasibility of Multi-Application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven IAEA Member States have been participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiment. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena including the coupling of primary system, high pressure containment and cooling pool are expected to occur in this transient. The ICSP has been conducted in three phases: pre-test (with designed initial & boundary conditions before the conduction of the experiment), blind (with real initial & boundary conditions after the conduction of the experiment) and open simulation (after the observation of real experimental data). Most advanced thermal-hydraulic system analysis codes like TRACE, RELAP5-3D and MARS have been assessed against experiments conducted at MASLWR test facility.

A Conceptual Design of a Small Modular Natural-Circulation Liquid Metal Fast Reactor With AMTEC Units

2013 ◽  
Author(s):  
Xun Zhang ◽  
Daogang Lu
Keyword(s):  
Liquid Metal ◽  
Conceptual Design ◽  
Fast Reactor ◽  
Tube Bundle ◽  
Natural Circulation ◽  
Power Conversion ◽  
Heat Removal ◽  
Design Parameters ◽  
Primary Loop ◽  
Conversion System

A conceptual design of a low power liquid metal fast reactor with Alkali-metal thermal-to-electric conversion (AMTEC) units is presented. The small modular reactor (SMR) can operate in remote locations as a distributed energy system. This paper provides the layout of the main components of the plant and main design parameters. The reactor core power of 2MW (t) may be completely removed by natural circulation, thus there is no need for primary pumps. Passive heat removal system (PHRS) is adopted to remove the decay heat. In addition, it could also be activated when part of the energy conversion units are disabled. AMTEC unit is a good candidate for converting the reactor power to electricity. An expanded AMTEC tube bundle system and the layout of power conversion system are both presented. The converters are coupled to the primary loop through six independent secondary loops. Liquid-liquid intermediate heat exchangers (IHXs) would be used to connect the primary loop and secondary loops as well as power conversion system and secondary loops. The design objectives for SMR-AMTEC system are to provide a greatly simplified power system with respect to design, construction, operation and maintenance.

Direct Contact Heat Exchange Interfacial Phenomena for Liquid Metal Reactors: Part II — Void Fraction

2002 ◽  
Author(s):  
S. Abdulla ◽  
X. Liu ◽  
M. H. Anderson ◽  
R. Bonazza ◽  
M. L. Corradini ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Exchange ◽  
Liquid Metal ◽  
Void Fraction ◽  
Direct Contact ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Test Facility ◽  
Contact Heat ◽  
Contact Heat Exchange

One concept being considered for steam generation in innovative nuclear reactor applications, involves water coming into direct contact with a circulating molten metal. The vigorous agitation of the two fluids, the direct liquid-liquid contact and the consequent large interfacial area can give rise to large heat transfer coefficients and rapid steam generation. For an optimum design of such direct contact heat exchange and vaporization systems, detailed knowledge is necessary of the various flow regimes, interfacial transport phenomena, heat transfer and operational stability. In order to investigate the interfacial transport phenomena, heat transfer and operational stability of direct liquid-liquid contact, a series of experiments are being performed in a 1-d test facility at Argonne National Laboratory and a 2-d experimental facility at UW-Madison. Each of the experimental facilities primarily consist of a liquid-metal melt chamber, heated test section (10cm diameter tube for 1-d facility and 10cm × 50cm rectangle for 2-d facility), water injection system and steam suppression tank. This paper is part II which, primarily addresses results and analysis of a set of preliminary experiments and void fraction measurements conducted in the 2-d facility at UW-Madison, part I deals with the heat transfer in the 1-d test facility at Argonne National Laboratory. A real-time high energy X-ray imaging system was developed and utilized to visualize the multiphase flow and measure line-average local void fractions, time-dependent void fraction distribution as well as estimates of the vapor bubble sizes and velocities. These measurements allowed us to determine the volumetric heat transfer coefficient and gain insight into the local heat transfer mechanisms. In this study, the images were captured at frame rates of 100 fps with spatial resolution of about 7mm with a full-field view of a 15cm square and five different positions along the test section height. The full-field average void fraction increases rapidly to about 15% in these preliminary tests, with the apparent boiling length of less than 20cm. The volumetric heat transfer coefficient between the liquid metal and water are compared to the CRIEPI data, the only prior data for direct contact heat exchange for these liquid metal/water systems.

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