Thermal-hydraulic Test Loop Research Articles

An analytical investigation of cold leg small break accidents of the ATLAS facility

A previous parametric study of the direct vessel injection (DVI) line breaks was re-evaluated to see its applicability to that of the cold leg (CL) pipe breaks in advanced thermal-hydraulic test loop for accident simulation (ATLAS). Evaluation results of the tests and analyses for the major parameters, e.g., the pressurizer (PZR) pressure, downcomer water level, collapsed core water level, and clad temperature, were compared for four different CL pipe break scenarios. The overall trends of the major parameters showed reasonable behaviors between the tests and analyses. The clad temperature showed conservative behaviors in the analyses using the suggested options. The suggested counter-current flow limit (CCFL) options for the fuel alignment plate (FAP) and cross-over legs (COLs) can be applicable to any small-break loss-of-coolant accident (SBLOCA) scenario for the CL pipe and DVI line breaks in the ATLAS tests.

Neutron Tomography Using Mobile Neutron Generators for Assessment of Void Distributions in Thermal Hydraulic Test Loops

Detailed knowledge of the lateral distribution of steam (void) and water in a nuclear fuel assembly is of great value for nuclear reactor operators and fuel manufacturers, with consequences for both reactor safety and economy of operation. Therefore, nuclear relevant two-phase flows are being studied at dedicated thermal-hydraulic test loop, using two-phase flow systems ranging from simplified geometries such as heated circular pipes to full scale mock-ups of nuclear fuel assemblies.Neutron tomography (NT) has been suggested for assessment of the lateral distribution of steam and water in such test loops, motivated by a good ability of neutrons to penetrate the metallic structures of metal pipes and nuclear fuel rod mock-ups, as compared to e.g. conventional X-rays, while the liquid water simultaneously gives comparatively good contrast. However, these stationary test loops require the measurement setup to be mobile, which is often not the case for NT setups. Here, it is acknowledged that fast neutrons of 14 MeV from mobile neutron generators constitute a viable option for a mobile NT system.We present details of the development of neutron tomography for this purpose at the division of Applied Nuclear Physics at Uppsala University. Our concept contains a portable neutron generator, exploiting the fusion reaction of deuterium and tritium, and a detector with plastic scintillator elements designed to achieveadequate spatial and energy resolution, all mounted in a light-weight frame without collimators or bulky moderation to allow for a mobile instrument that can be moved about the stationary thermal hydraulic test sections. The detector system stores event-to-event pulse-height information to allow for discrimination based on the energy deposition in the scintillator elements. Experimentalresultsfrom the tomographic assessment of axially symmetric test objectsare shown, as well as simulation results from a scaled up version of the instrument for nonsymmetrical objects in quarter fuel-bundle size objects. In conclusion, the application of tomography on inch-wide vertical pipes has been experimentally demonstrated and simulation results indicate that tomography of the void distribution in nonsymmetrical vertical flows in quarter BWR fuel bundles is also feasible.

Thermal–hydraulic analysis and code assessment for ATLAS 6-inch cold leg break (SBLOCA) test using MARS-KS

The thermal–hydraulic analysis using MARS-KS code was performed for 6-inch cold leg break test of ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation), which was the second domestic standard problem. The calculation results were compared with experimental data to assess the code capability to simulate the transient thermal–hydraulic behavior for small break loss of coolant accident (SBLOCA). The sequence of events, except for the location of loop seal clearing (LSC) and safety injection tank (SIT) injection time was predicted well. The loop seals of 1A and 2B intermediate legs were cleared at 398s in the experiment, while that of 1A was only cleared in the calculation at the same time. The prediction showed good agreement with the experimental data for pressurizer pressure and break mass flow rate. The sudden decrease and increase of water level at the LSC time were predicted qualitatively. After LSC, there was significant water level dip at SIT injection time which was not seen in the experiment. In addition, sensitivity study to investigate the cause of core level dip at SIT injection time was performed and discussions were made for it. In conclusion, MARS-KS code has good capabilities to simulate cold leg break SBLOCA, however, including interfacial heat and mass transfer, especially condensation model needs to be improved to predict more accurate results.

Development of a phenomena identification ranking table for simulating a station blackout transient of a pressurized water reactor with a thermal-hydraulic integral effect test facility

KAERI (Korea Atomic Energy Research Institute) will launch an OECD/NEA project by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from April, 2014. In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to identify the major thermal-hydraulic phenomena of high risk multiple failure accidents considering magnified safety concerns after the Fukushima accident. A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project.In this study, a phenomena identification ranking table (PIRT) has been developed for identifying the major parameters affecting the thermal-hydraulic phenomena in SBO transients. Development of a PIRT for the SBO transients is expected to contribute to making strategy for performing a thermal-hydraulic integral effect test with ATLAS which includes determination of proper test conditions and improvement of measurements. The primary safety criterion (PSC) was determined to be the core mixture level from the expert panel discussion. As for the major components of the nuclear power plant, the rank of the importance and the knowledge level were summarized for each of the three temporal phases. Taking into account the important phenomena from the PIRT result, it can be confirmed that ATLAS has a capability of simulating the major thermal-hydraulic phenomena during a SBO transient as realistically as possible. The present PIRT will be directly used to make strategy for performing a thermal-hydraulic integral effect test with ATLAS in the framework of the OECD-ATLAS project.

Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator.

In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the time-averaged void distribution in these loops. Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantage of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed. In this article, three axially symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm, and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper. Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020, and 0.022 cm(-1), solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful indication on the flow mode, and a visualization of the radial material distribution can be obtained. A benefit of this system is its potential to be mounted at any axial height of a two-phase test section without requirements for pre-fabricated entrances or windows. This could mean a significant increase in flexibility of the void distribution assessment capability at many existing two-phase test loops.

SECOND ATLAS DOMESTIC STANDARD PROBLEM (DSP-02) FOR A CODE ASSESSMENT

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This 2nd ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

Analysis of an ATLAS 6-in. cold-leg break simulation with MARS code

A transient analysis of an ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) 6-in. cold-leg break was carried out with the MARS (Multi-dimensional Analysis for Reactor Safety) safety analysis code in the frame of the Domestic Standard Problem exercise in Korea. The calculated major sequence of events of the 6-in. cold-leg break simulation showed good agreement with the measured data. The calculated break mass flowrate was predicted well, whereas accumulated mass of the break outflow was underestimated due to the underestimation of the break mass flowrate in a later phase. The general trends of the collapsed water level were well predicted in the core and the downcomer region. The loop seal clearing phenomena were observed at about 400s in the 1-A and 2-B intermediate legs in the calculation results which is identical to the experiment. To investigate the emergency core coolant (ECC) bypass phenomena, variation of the boron concentration in safety injection water was adopted. In the loop seal clearing phase of the 6-in. cold-leg break, about 40–50% of the ECC bypass fraction was calculated and after the loop seal clearance, 20–30% of ECC water bypassed, that is, not participating in core cooling directly. The boron tracking approach is considered to be a feasible methodology with which quantify the ECC bypass flow in the early phase of a small-break loss-of-coolant accident (SBLOCA).

A SUMMARY OF 50thOECD/NEA/CSNI INTERNATIONAL STANDARD PROBLEM EXERCISE (ISP-50)

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.

Integral Effect Tests on Transient Thermal-Hydraulic Behavior during a Steam Generator Tube Rupture Accident in the APR1400

A postulated steam generator tube rupture (SGTR) event of the APR1400 (Advanced Power Reactor 1400 MWe) was experimentally investigated with the thermal-hydraulic integral effect test facility ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation). The SGTR accident is one of the design-basis accidents having a significant impact on safety from the viewpoint of radiological release. To simulate a SGTR accident of the APR1400, the SGTR-HL-04 and the SGTR-HL-05 tests were performed by simulating double-ended ruptures of a single U-tube and five U-tubes at the hot side of the ATLAS steam generator. Following the reactor trip induced by a high steam generator level signal, the primary-system pressure decreased and the secondary-system pressure increased until the main steam safety valves were opened to reduce the secondary-system pressure. A mild change of the water level in the core was observed, which was attributed to the small break sizes of the present tests compared with conventional loss-of-coolant-accident tests. No excursion in the cladding temperature was observed in either test. The break area affected the timing of the major events observed in the tests. Lessened heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. The break flow was discharged as single-phase water in both tests.

Major Achievements and Prospect of the ATLAS Integral Effect Tests

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

Effects of Break Size on Direct Vessel Injection Line Break Accidents of the ATLAS

The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break among the analyzed small-break loss-of-coolant accidents in safety analysis. The first-ever integral effects test database for various DVI line break sizes from 5% to 100% was established with the Korea Atomic Energy Research Institute’s Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) test facility. This database enhances our physical understanding of the major thermal-hydraulic behaviors of the APR1400 during DVI line break accidents, and it can also be used to examine the prediction capabilities and identify any deficiencies in the existing best-estimate safety analysis codes. Effects of the break size were experimentally investigated, and the best-estimated MARS code was assessed against the experimental database. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than the data just before a loop seal clearing occurs, and it also produced a more rapid decrease in the downcomer water level than the data. These disagreements are the expected consequence of uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effects test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology on the DVI line break accidents of the APR1400.

FIRST ATLAS DOMESTIC STANDARD PROBLEM (DSP-01) FOR THE CODE ASSESSMENT

KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basis accidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed and successfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from the ATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. For the first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 was selected by considering its technical importance and by incorporating comments from participants. Twelve domestic organizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. This ATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participants prior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was also used by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison results between the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations for code users as well as for developers are suggested.

A separate effect test on the cooling behavior in a pressurized water reactor core under a low reflooding rate condition by using ATLAS and RELAP5/MOD3.3 code assessment

A separate effect test was performed on the cooling behavior in a PWR core under a low reflooding rate condition by using the ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) which is a thermal–hydraulic integral effect test facility for the pressurized water reactors APR1400 and OPR1000. Although several integral tests for the reflood phase of a large break loss of coolant accident (LBLOCA) of APR1400 have been performed with the ATLAS, the previous integral effect tests for the reflood phase of a LBLOCA are not easily simulated by existing codes, such as the RELAP5/MOD3, due to a unique phenomena in ATLAS, that resulted from an injection of large amount of subcooled water onto the heated wall of which temperature was higher than the target value.Therefore, there is a need to perform a separate effect test under a low reflooding rate condition using the ATLAS to help validate the RELAP5 reflood models. The present LB-CL-15 test was performed to validate the RELAP5 reflood models for a core quench phenomenon under a low flow rate ECC injection condition. Separate effect test data for APR1400 could be obtained by using the ATLAS, which could simulate a direct vessel injection, a reverse heat transfer from steam generators and the steam binding effect, etc. The experimental data were used to assess the reflood models of the safety analysis codes of the original RELAP5/MOD3.3 and a modified version of RELAP5/MOD3.3. The modified version showed better agreement with the test data than the original version.

An Integral Effect Test on the Reflood Period of a Large-Break LOCA for the APR1400 Using ATLAS

A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), has been constructed at the Korea Atomic Energy Research Institute. It is a 1/2-reduced-height and 1/288-volume-scaled test facility based on the design features of APR1400, an evolutionary pressurized water reactor developed by the Korean industry. ATLAS was used to perform a set of integral effect tests on the reflood period of a large-break loss-of-coolant accident (LBLOCA) after intensive performance tests had been conducted to verify ATLAS’s operational performance and controllability for major thermal-hydraulic components. The present LB-CL-09 test is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period that can be used to provide reliable data to help validate the LBLOCA analysis methodology for APR1400. The main objective of the present test is to identify the major thermal-hydraulic characteristics such as the direct emergency core coolant (ECC) bypass, downcomer boiling, and core cooling behavior during the reflood phase of an LBLOCA for APR1400 under conditions where the downcomer region interacts with the reactor core region and the heat could be transferred through the steam generator. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results. The decay heat and the ECC flow rate from the safety injection tank were simulated from the start of the reflood period. The ECC flow rate from the safety injection pump was 0.32 kg/s. The system pressure was fixed at ~0.1 MPa, and the initial outer-wall temperature was determined to be 205°C. The experimental results showed the typical thermal-hydraulic trends expected to occur during the reflood phase of the LBLOCA scenario.

Major findings from LBLOCA reflood tests using the ATLAS facility

A set of LBLOCA (large-break loss of coolant accident) reflood tests was performed in the first phase of the ATLAS (advanced thermal-hydraulic test loop for accident simulation) program. Their main objectives were to identify the major thermal-hydraulic characteristics during the reflood phase of a LBLOCA for APR1400 and to provide qualified data for APR1400 licensing. The ATLAS reflood test program could be divided into two phases (Phase-1 and Phase-2) according to the target period to be simulated. The Phase-1 tests were parametric effect tests for downcomer boiling in the late reflood phase of LBLOCA and the Phase-2 tests were integral effect tests for the entire reflood phase of LBLOCA. The experimental results from both Phase-1 and Phase-2 tests reproduced typical thermal-hydraulic trends expected to occur during the APR1400 LBLOCA scenario. A separate effect test was also performed under a low reflooding rate condition to provide data to validate the RELAP5 reflood models, and its experimental results showed a gradual reflooding in the core, a subsequent quenching of the core heater rods and the cooling of the reactor pressure vessel downcomer.

An Integral Effect Test on the LBLOCA Reflood Phenomena for the APR1400 Using ATLASunder a Best-Estimate Condition

The Korea Atomic Energy Research Institute (KAERI) had started the operation of the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors of APR1400 and OPR1000. Recently, integral tests for the reflood phase of a large-break loss of coolant accident (LBLOCA) have been performed with the ATLAS after an extensive series of commissioning tests. The reflood tests include both Phase-1 and Phase-2 tests. Phase-1 tests are parametric effect tests for the downcomer boiling phenomena in a reactor pressure vessel during the LBLOCA late reflood period, and Phase-2 tests are integral effect tests for the thermal-hydraulic phenomena in the core and downcomer during the LBLOCA reflood period to provide important thermal-hydraulic parameters such as the peak cladding temperature for the evaluation of the safety analysis code and the corresponding licensing methodology. In the present paper, the LB-CL-14 test, which is an integral effect test on the LBLOCA reflood phenomena for APR1400 using the ATLAS under a best-estimate condition, is analyzed and the major findings are presented. As a typical radial power profile is given to the core heater and the system pressure varies to simulate the APR1400 as realistically as possible, the present test gives the most realistic results among the Phase-2 tests in terms of the peak cladding temperature (PCT) and rewetting time.

EXPERIMENTAL SIMULATION OF A DIRECT VESSEL INJECTION LINE BREAK OF THE APR1400 WITH THE ATLAS

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.

An Integral Effect Test on the LBLOCA Reflood Phenomena for the APR1400 Using ATLAS under a Best-Estimate Condition

The Korea Atomic Energy Research Institute (KAERI) had started the operation of the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors of APR1400 and OPR1000. Recently, integral tests for the reflood phase of a large-break loss of coolant accident (LBLOCA) have been performed with the ATLAS after an extensive series of commissioning tests. The reflood tests include both Phase-1 and Phase-2 tests. Phase-1 tests are parametric effect tests for the downcomer boiling phenomena in a reactor pressure vessel during the LBLOCA late reflood period, and Phase-2 tests are integral effect tests for the thermal-hydraulic phenomena in the core and downcomer during the LBLOCA reflood period to provide important thermal-hydraulic parameters such as the peak cladding temperature for the evaluation of the safety analysis code and the corresponding licensing methodology. In the present paper, the LB-CL-14 test, which is an integral effect test on the LBLOCA reflood phenomena for APR1400 using the ATLAS under a best-estimate condition, is analyzed and the major findings are presented. As a typical radial power profile is given to the core heater and the system pressure varies to simulate the APR1400 as realistically as possible, the present test gives the most realistic results among the Phase-2 tests in terms of the peak cladding temperature (PCT) and rewetting time.

Simplified numerical model for defining Ledinegg flow instability margins in MTR research reactor

Establishment of safety margins and the corresponding operating condition limits will ensure achievement of a safe operation of nuclear installations. For this purpose, several critical phenomena have been analyzed theoretically and experimentally and a great number of models and correlations are made available. Among these critical issues the well-known flow instability has been intensively investigated by several authors especially for nuclear power plants' (NPPs) operating conditions. However, limited published work is available for research reactor operation conditions. In general, the Whittle and Forgan correlation is widely used to define the margin to static flow instabilities in narrow parallel heated channels for research reactors. In the framework of verification and assessment of the capabilities of the RELAP5/Mod 3 system code to determine the onset of flow instability in research reactor conditions, a simple model based on steady-state equations adjusted with drift-flux correlations has been developed. The program is used to draw the pressure drop characteristic curves and to establish the conditions of the Ledinegg instability in a uniformly heated channel subject to constant outlet pressure. The model is assessed by using experimental data from a thermal hydraulic test loop by Siman-Tov and numerical results from RELAP5/Mod 3. The model presents acceptable estimation of the target mass flow that would induce flow instability and the latter could be then used to establish a conservative margin to the Ledinegg instability.

Development of a pump performance model for an integral effect test facility

Modeling methods for the reactor coolant pumps of the existing integral facilities are reviewed from the viewpoint of scaling, single- and two-phase characteristics. A series of separate effect tests was performed for the pumps of the ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) in order to obtain a complete set of single-phase homologous curves in all the quadrant operating regions. Besides the friction loss tests for the sheared shaft and the locked rotor conditions were conducted to extend the homologous curves to the limiting cases. A method which can take into account the two-phase degradation effects of the ATLAS pumps was suggested as a first approximation based on similarity principles. The present data and model can completely replace the pump input model of the MARS 3.1 code both for single- and two-phase conditions.