Shutdown Heat Removal Test Research Articles

Validation of Pronghorn’s subchannel code using EBR-II shutdown heat removal tests: SHRT-17 and SHRT-45R

Pronghorn-Subchannel, referred to as Pronghorn-SC throughout this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Pronghorn-SC was initially designed to model flows in water-cooled, square-lattice, bare subassemblies. Its capability has been extended to model flows in liquid-metal-cooled, triangular-lattice, wire-wrapped subassemblies. To ensure the accuracy of Pronghorn-SC in predicting the behavior of liquid sodium-cooled reactors, the code was validated by comparing calculations with experimental data, obtained from the experimental breeder reactor (EBR-II), Shutdown Heat Removal Tests (SHRT) 17 and 45R. The steady-state calculation at the beginning of the transients was validated using temperature measurements taken at different axial elevations in the instrumented subassembly XX09. The validation exercise was performed in successive stages. First, a comparison between the measured temperature profiles and standalone Pronghorn-SC simulations using a uniform pin power profile was made. The pin power profile was then refined using a Serpent-2 model of the reactor core. Finally, the radial temperature profile was further corrected considering the inter-assembly heat transfer. A Pronghorn Finite-Volume (FV) thermal hydraulic simulation of XX09 and its six neighboring subassemblies; calculated the heat flux on the inner duct surface of the XX09 subassembly to inform the Pronghorn-SC model. Last, a transient validation calculated the peak temperature evolution during the SHRT tests.

Effect of inlet throttling on thermohydraulic instability in a large scale water-based RCCS: An experimental study

The objective of the present experimental study is to investigate the effect of inlet throttling on the thermohydraulic stability of a large scale water-based Reactor Cavity Cooling System (RCCS). The test was performed using the water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne, which represented a ½ axial scale and 12.5° sector slice of the full scale Framatome 625 MWt SC-HTGR RCCS concept. A two-phase steady state was first established through direct condensate refill, followed by increased inlet throttling over 10 stages, corresponding to a loss coefficient K over the range of 0.05–653. With the inlet throttling gradually increased, the system experienced a unique transition process between stabilization and destabilization. Through a stability analysis, three instability mechanisms were identified in the present test, including a compound mechanism due to both natural circulation oscillations (NCOs) and density wave oscillations (DWOs), Type-II DWOs, and geysering.

EXTENDING CTF MODELING CAPABILITIES TO SFRs AND VALIDATION AGAINST SHRT TESTS

The utilization of liquid metals as coolants for fast reactors brings several economical and practical advantages that lead to a sustainable future for nuclear energy. Molten sodium is used as a coolant in Sodium Fast Reactors (SFRs). Sodium is relatively cheaper than other metal coolants. It requires lower pumping power, causes less neutron moderation and it is non-corrosive to the fuel cladding. The SFR hexagonal subassemblies are relatively smaller than Light Water Reactors (LWRs) subassemblies. The differences in the geometrical design of SFRs compared to LWRs lead to different physical behavior of the coolant. Several models and correlations particular to sodium were implemented in thermal-hydraulics (TH) computer codes in order to model the coolant behavior accurately. CTF is a subchannel TH code that was designed and validated for LWRs. In this work, the capabilities of the code were extended to SFRs by incorporating sodium coolant properties and correlations for friction factor, flow mixing coefficient and conduction heat transfer. The code was then validated against selected steady state data from the Experimental Breeder Reactor II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R. CTF was used to simulate the instrumented subassemblies XX09 and XX10. The results demonstrate the capability of CTF to model SFRs. Code validation is currently being extended to the transient phases of the SHRT experiments.

Benchmark Simulation of the Natural Convection Shutdown Heat Removal Test Facility Using SAM

Natural convection systems are a promising method to passively remove heat from reactor cavities during loss of forced flow accident scenarios. At Argonne National Laboratory (ANL), a highly instrumented Natural Convection Shutdown Heat Removal Test Facility (NSTF) was used to demonstrate the effectiveness of air-cooled natural convection systems. In previous work, RELAP5-3D simulations were performed on this facility with favorable comparisons to experiment for mass flow rate, pressure drop, air temperature increase, and air velocity. Both experimental and simulation efforts with this facility present a useful opportunity to perform a benchmark study with the System Analysis Module (SAM). SAM is an advanced thermal-hydraulic system code currently in development at ANL for advanced non–light water reactor safety analysis.

Benchmark of GOTHIC to EBR-II SHRT-17 and SHRT-45R Tests

GOTHIC™ has been used to simulate the Experimental Breeder Reactor–II (EBR-II) Shutdown Heat Removal Test 17 (SHRT-17) and Shutdown Heat Removal Test 45R (SHRT-45R), which correspond to protected and unprotected loss-of-flow events, respectively. GOTHIC is a versatile general-purpose, thermal-hydraulic software package that is a hybrid between traditional system thermal-hydraulic and computational fluid dynamics codes. It is a practical engineering tool that has been used for the design and licensing of existing plants, small modular reactors (SMRs), and next-generation plant designs. Historically, the software has been applied for containment analysis and operability assessments for light water reactors (LWRs), but the recent improvements included in GOTHIC 8.3(QA) allow for the software to be used to simulate advanced, non-LWR concepts currently being developed such as sodium, molten salt, lead, and gas–cooled designs. It will be demonstrated in this paper that GOTHIC includes both the required attributes to model EBR-II and the appropriate physics to accurately simulate the steady-state operating conditions as well as SHRT-17 and SHRT-45R. The GOTHIC model of EBR-II was developed using only publicly available information. The nodalization was selected not only to capture the important phenomena but also to remain computationally efficient. The GOTHIC results show good agreement in both magnitude and trend with the experimental data. Differences are within the bounds of experimental uncertainty and required engineering assumptions applied in the model to fill in gaps in information, particularly for the various leakage paths that existed throughout the primary side of EBR-II, and were not well characterized during the tests.

Development and preliminary verification of code IMPC-transient

A project of Accelerator Driven Advanced Nuclear Energy System (ADANES) has been proposed in China. The system safety analysis code IMPC-transient was developed to provide suitable tools for the safety analysis and scheme optimization of the sub-critical reactor in ADANES. In IMPC-transient, the thermal-hydraulic (TH) module is developed independently. In every time step, the momentum equation and the energy equation are one-way coupled, and the system energy equation is divided into different nodal volume. The TH module was then internally coupled with the deterministic neutronic code DAISY, in which both the neutron transport kinetics and the point reactor kinetics are available. On the steady condition, The thermal-hydraulic module and neutronic (NE) module are two-way coupled. On transient conditions, TH module and NE module are one-way coupled. The sodium was added into IMPC-transient material package and the benchmark “EBR-II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R” was used for the verification of the code. The results show that IMPC-transient is reliable and valid.

Ambient and building condition effects modeling of air-cooled natural circulation systems

Throughout the testing program at Natural convection Shutdown heat removal Test Facility (NSTF), strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under identical test procedures. Significant analysis efforts were conducted to gain understanding of these influences, and quantify the system response of the test facility. An empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions, including wind and temperatures. Coefficients in the model were determined based on best fitting of the experimental data set. The predictive capability of the empirical model was demonstrated by applying it to new sets of experimental data. After implementing the empirical model in both system-level thermal fluids and computational fluid dynamics modeling of the facility, simulations from both approaches resulted in accurate predictions in system natural circulation flow rates.

Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R – (2) subchannel analysis of instrumented fuel subassembly

Two kinds of Loss of Flow experiments conducted in EBR-II are the subject of an IAEA benchmark exercise. The present work discusses the subchannel analysis in an instrumented subassembly, analyzed with the COBRA-IV-I code. Boundary conditions are provided by a 1D thermal–hydraulic system code (NETFLOW++) for a protected loss-of-flow and an unprotected-loss-of-flow tests conducted at the EBR-II reactor. The instrumented subassembly with 61 pins is installed in the 5th row of the reactor core in order to measure the temperature profile across the mid-plane and the top region of the fuel subassembly. The pin lattice in the subassembly consists of 126 subchannels. Calculated temperature profiles at three levels are compared with the measured results. Biased temperature profiles with respect to the center line of the subassembly measured in the experiments are appropriately simulated by the code. The presence of the spacer wire by itself introduces a biased temperature profile. The biased temperature profile is also shown by a CFD calculation using a 7-pin partial model of the instrumented subassembly. In order to obtain close agreement with measured temperatures, the power profile within the subassembly must be taken into account. Temperature evolutions during the transients are simulated by the subchannel code.

Assessment calculation of MARS-LMR using EBR-II SHRT-45R

KAERI has designed a prototype Gen-IV SFR (PGSFR) with metallic fuel. And the safety analysis code for the PGSFR, MARS-LMR, is based on the MARS code, and supplemented with various liquid metal related features including sodium properties, heat transfer, pressure drop, and reactivity feedback models. In order to validate the newly developed MARS-LMR, KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (SHRT)”. Argonne National Laboratory (ANL) has technically supported and participated in this program. One of benchmark analysis tests is SHRT-45R, which is an unprotected loss of flow test in an EBR-II. So, sodium natural circulation and reactivity feedbacks are major phenomena of interest. A benchmark analysis was conducted using MARS-LMR with original input data provided by ANL. MARS-LMR well predicts the core flow and power change by reactivity feedbacks in the core. Except the results of the XX10, the temperature and flow in the XX09 agreed well with the experiments. Moreover, sensitivity tests were carried out for a decay heat model, reactivity feedback model, inter-subassembly heat transfer, internal heat structures and so on, to evaluate their sensitivity and get a better prediction. The decay heat model of ANS-94 shows better results of fission power, however, the fission power is still over-estimated in the long-term transient region by the reactivity feedbacks. The inter-subassembly heat transfer is the most influential parameter, especially for the non-fueled XX10, which has a low flow and power subassembly. In addition, the appropriate internal heat structure model can be an influential parameter. Finally, the corrected results are proposed with reasonably conjectured parameters. This study can give the validation data for the MARS-LMR and better understanding of the EBR-II SHRT-45R.

Full-core coupled neutronic/thermal-hydraulic modelling of the EBR-II SHRT-45R transient

During the last decade the European activities in the field of nuclear fission research include the design of fast reactors cooled by liquid metals. Within this framework, the Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code is being developed at Politecnico di Torino over the last few years. It implements a full-core coupled neutronic/thermalhydraulic model of a liquid-metal-cooled fast reactor as relevant for two of the six options currently under study within the framework of the Generation-IV International Forum, namely the lead-cooled fast reactors and the sodium-cooled fast reactors. The code validation process involves the participation in a coordinated research project of the International Atomic Energy Agency, aiming at testing different computational tools against the shutdown heat removal tests performed many years ago in the sodium-cooled Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory, USA. In this paper, results of the FRENETIC analysis of one of the transients considered in the project, the unprotected EBR-II shutdown heat removal test SHRT-45R, are presented and compared to the measurements, providing the first validation of the coupled neutronic/thermal-hydraulic features of the FRENETIC code.

Performance test of MARS-LMR code with benchmark analysis of EBR-II SHRT-17

The MARS-LMR code has been developed to analyze transients in a pool-type sodium-cooled fast reactor (SFR) which has been designed by the Korea Atomic Energy Research Institute (KAERI). Currently, KAERI has developed a prototype Gen-IV SFR (PGSFR) with metallic fuel. The MARS-LMR code is based on the MARS code, and supplements various liquid metal related features including sodium properties, sodium heat transfer and pressure drop models. In order to validate the newly developed MARS-LMR, KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on the “Benchmark Analysis of an EBR-II shutdown heat removal test” (Briggs et al., 2013, 2015). Argonne National Laboratory (ANL) has operated the EBR-II technically supported and is participating in this program (Sumner and Wei, 2012). In addition, various institutes and countries have joined the IAEA-CRP. The EBR-II SHRT-17 is a kind of loss of flow test, and thus sodium natural circulation is a major phenomenon. The benchmark analysis is conducted using MARS-LMR with the original input data provided by ANL. Qualitatively, MARS-LMR predicted well the natural circulation flow in the core. In addition, sensitivity tests were carried out for various influential parameters to get a better prediction. This study can give validation data for MARS-LMR and a better understanding of EBR-II SHRT-17.

Experimental observations of natural circulation flow in the NSTF

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary of some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states.

Validation of a Three-Dimensional Model of EBR-II and Assessment of RELAP5-3D Based on SHRT-17 Test

The International Atomic Energy Agency established a Coordinated Research Project (CRP) for EBR-II shutdown heat removal tests (SHRT). The CRP aims at improving the design and the simulation capabilities in fast reactor neutronics, thermal hydraulics, plant dynamics, and safety analyses. This is achieved by benchmark analyses of protected (SHRT-17) and unprotected (SHRT-45r) loss-of-flow tests, from the EBR-II SHRT program. In this framework, ENEA has set up, applied, and is validating an integrated multiphysics approach, based on existing codes, for supporting the design and the safety analysis of Generation IV liquid-metal fast reactors. This paper outlines the rationale of the CRP participation, and it focuses on the qualification of a three-dimensional (3-D) thermal-hydraulic nodalization of EBR-II and on the assessment of RELAP5-3D code against the test SHRT-17. The nodalization models one by one the fuel assemblies of the core and of the extended core of the reactor for an efficient coupling with a 3-D neutron kinetic analysis code. The experimental data are presented and the thermal-hydraulic phenomena of test SHRT-17 are discussed, being the basis for assessing the code performance and for discussing its limitations. Blind and open calculation results are presented and discussed.

Thermal-hydraulic analysis of EBR-II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R

As one of the six Generation-IV reactors, SFR (sodium-cooled fast reactor) is attracting more and more attention in recent years. To evaluate the SFR safety characteristics, the Shutdown Heat Removal Tests (SHRT) conducted by Argonne National Laboratory (ANL) were performed in Experimental Breeder Reactor-II (EBR-II) to verify inhere safety characteristics. To find the inaccuracies in the exist models and improve the simulating level for SFR, SHRT-17 and SHRT-45R have been chosen as the benchmark problems for the IAEA coordinated research project (CRP), which we have participated in. THACS (the Transient Thermal-Hydraulic Code for Analysis of Sodium Cooled Fast Reactor) developed by XJTU was used in our work. THACS has finished preliminary with completion of all the models of primary loop originally intended for simulating accidents in SFRs. THACS is a multi-channel, multi-phase, multi-component transient and universal code coupled with the point reactor kinetics model. In this paper the models of THACS were introduced in detail and SHRT scenario simulation work of EBR-II has been performed. THACS was verified with the simulation results of flowrate and temperatures in core and instrumented subassembly XX09. After the verification, the peak coolant, cladding and fuel temperatures were estimated which is under the safety limits. Results show that EBR-II is effective in protecting the reactor against the severe loss-of-flow accidents. The three-layer model for cold pool coupled with THACS was used and the predicted results in a better agreement with the measurements.

Basic verification of THACS for sodium-cooled fast reactor system analysis

For the transient thermal–hydraulic analysis of sodium cooled fast reactor (SFR), a universal code THACS is being developed in China. The one-dimensional modules for the main components of the primary loop, intermediate loop and heat removal loop have been compiled and the models are introduced briefly in this paper. For two-phase calculation, the multi-bubble model is used in the core module for sodium, while a simple incompressible water model is applied in the water side of steam generator. The point kinetic equations are employed to calculate the fission power of the multi-channel core, and various kinds of feedbacks like Doppler Effect, radial and axial expansions, density effect are taken into account. Preliminary verifications of the code are done based on the benchmark analyses of a BN-800 type SFR under loss of flow (LOF) accident, the EBR-II shutdown heat removal tests, and the 1-D simulation of the out-of-pile LOF experiment in KNS. Code-to-code and code-to-experiment comparisons show that THACS is proper for the transient analysis of SFR.

B124 高速炉EBR-IIの自然循環試験IAEA国際ベンチマーク解析研究(OS3 軽水炉・新型炉・原子力安全(2))

A Shutdown Heat Removal Test (SHRT) conducted using the EBR-II reactor at Argonne National Laboratory (ANL) in the U.S.A. was proposed by International Atomic Energy Agency (IAEA) in June 2012 as a benchmark problem. The SHRT-17 test is loss-of-flow (LOF) test, and the SHRT-45R is an unprotected loss-of-flow (ULOF) test. We use the plant system code NETFLOW++ and the neutronics code ERANOS to analyze the two tests. The SHRT-17 test is calculated using only the NETFLOW++ code because it is a purely thermal hydraulic problem. To calculate the SHRT-45R test, we should have delayed neutron data and reactivity coefficients calculated by the ERANOS code. These parameters are used in the NETFLOW++ code to conduct a neutronics - thermal hydraulic coupled calculation. In the present work, major parameters of the plant are predicted with good accuracy. But we overestimated the peak temperature of coolant because of the slightly weak reactivity feedback coefficient for sodium density change in the fuel region. When the reactivity feedback coefficient is adjusted, good agreement of the peak coolant temperature is obtained.

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

Verification and Implications of the Multiple-Pin Treatment in the SASSYS-1 Liquid-Metal Reactor Systems Analysis Code

As part of a program to obtain realistic, as opposed to excessively conservative, analysis of reactor transients, a multiple-pin treatment for the analysis of intrasubassembly thermal hydraulics has been included in the SASSYS-1 liquid-metal reactor systems analysis code. This new treatment has made possible a whole new level of verification for the code. The code can now predict the steady-state and transient responses of individual thermocouples within instrumented subassemblies in a reactor rather than just predicting average temperatures for a subassembly. Very good agreement has been achieved between code predictions and the experimental measurements of steady-state and transient temperatures and flow rates in the shutdown heat removal tests in the Experimental Breeder Reactor II (EBR-II). Detailed multiple-pin calculations for blanket subassemblies in the EBR-II demonstrate that the actual steady-state and transient peak temperatures in these subassemblies are significantly lower than those that would be calculated by simpler models.

Development of Scaling Requirements for Natural Convection Liquid-Metal Fast Breeder Reactor Shutdown Heat Removal Test Facilities

This paper summarizes similitude requirements for liquid-metal fast breeder reactor natural convection shutdown heat removal test facilities. It is generally not possible under natural circulation ...