Pradip Saha

Hot-channel stability of supercritical water-cooled reactors-I: Steady state and sliding pressure startup

The drastic change of fluid density in the reactor core of a supercritical water-cooled reactor (SCWR) gives rise to a concern about density-wave stability. Using a single-channel thermal-hydraulic model, stability boundary maps for the U.S. reference SCWR design have been constructed for both steady state and sliding pressure startup conditions. The supercritical water flow in the reactor core has been simulated using a three-region model: a heavy fluid with constant density, a mixture of heavy fluid and light fluid similar to a homogeneous-equilibrium two-phase mixture, and a light fluid, which behaves like an ideal gas or superheated steam. Two important nondimensional numbers, namely, a pseudosubcooling number Npsub and an expansion number Nexp, have been identified for the supercritical region. The stability map in the supercritical region is then plotted in the plane made of these two numbers. The U.S. reference SCWR design operates in a stable region with a large margin. Sensitivity studies produced results consistent with the findings of the earlier research done for the subcritical two-phase flow. During the sliding pressure startup of the SCWR, a two-phase steam-water mixture at subcritical pressure will appear in the reactor core. A nonhomogeneous (e.g., drift-flux) nonequilibrium two-phase flow model was applied. The characteristic equation was numerically integrated, and stability boundary maps were plotted on the traditional subcooling number versus phase change number (or Zuber number) plane. These maps have been used to develop a sliding pressure SCWR startup strategy avoiding thermal-hydraulic flow instabilities.

HOT-channel stability of supercritical water-cooled reactors-II: Effect of water rod heating and comparison with BWR stability

The single hot-channel thermal-hydraulic stability model is expanded to investigate the effects of heat transport from fuel rods and to water rods on supercritical water-cooled reactor (SCWR) stability. Furthermore, the stability margin of the SCWR is compared with that of a typical boiling water reactor (BWR) by conducting a sensitivity study on operating conditions. The fuel thermal-dynamic effect is studied by coupling a lumped-parameter fuel model with the three-region coolant thermal-hydraulics model. It is found that the fuel heat capacity would dampen the oscillations in the coolant channel and therefore increase the stability of the system. Also, heating of the water rods, which could be allowed in the core, would improve single-channel stability. The stability sensitivity to power and flow rate conditions is analyzed for the U.S. reference SCWR design and compared with a typical BWR. The SCWR is found to be more sensitive to power and flow rate changes than the typical BWR. The water rod heating cannot significantly improve this sensitivity feature of the SCWR stability. The traditional stability measure of oscillation amplitude decay ratio does not capture the extent to which a stability margin exists in a particular design of the SCWR. The robustness of stability should be ascertained by examining accommodation of the potential variation and/or uncertainty about the nominal conditions.

Core-Wide (In-Phase) Stability of Supercritical Water-Cooled Reactors - I: Sensitivity to Design and Operating Conditions

Using a three-region supercritical water flow model, the core-wide in-phase stability of the U.S. reference supercritical water-cooled reactor (SCWR) design is investigated. The reactor core is simulated as three channels according to the radial power distribution. A method based on λ modes (reactivity modes) expansion of neutronic kinetic equations is applied. A constant pressure drop boundary condition between the feedwater pump and the turbine control valve is assumed. Cases with and without water rods heating are studied. It is found that the stability of the U.S. reference SCWR design is sensitive to the flow restrictions in the hot fluid or the steam line. As long as the restriction in the steam line is small, the design will be stable. A pressure loss coefficient of 0.25 is assumed for the exit valve on the steam line in this analysis. With this value, the SCWR is stable with a large margin. It is concluded that the presence of water rods heating will reduce the stability margin and increase the flow rate sensitivity while maintaining the power sensitivity level. The decay ratios for the three density wave oscillation modes, i.e., single hot channel, coupled neutronic out-of-phase and in-phase, are compared at steady-state conditions. It is found that the single hot channel oscillation mode is the most limiting one in the absence of the water rods heating, while the in-phase oscillation mode is most limiting in the presence of water rods heating.

Safety analysis of high-power-density annular fuel for PWRs

This paper assesses the performance of internally and externally cooled annular fuel in a four-loop pressurized water reactor during a variety of transients and accidents, namely, the loss of flow accident (LOFA), main steam line break (MSLB), large break loss of coolant accident (LBLOCA), and rod ejection accident (REA). The RELAP5 code was the primary vehicle for these analyses, although the VIPRE code was also used to calculate the minimum departure from nucleate boiling ratio (MDNBR) for LOFA and MSLB transients based on the RELAP5 results. It has been found that the MDNBR for the annular fuel at 150% power was higher than the MDNBR value for the reference solid fuel at 100% power for LOFA and MSLB. For LBLOCA analysis, the RELAP5-3D code was applied twice since the code has a constraint on the reflood model, which can be applied to only one cooling surface (either the inner channel or the outer channel). The analysis, with the reflood model applied to the outer channel, showed that using the standard size (100%) accumulator but with an increased (150%) safety injection flow rate, the peak cladding temperature (PCT) for the annular fuel at 150% power would be ~1200 K (927°C). This is ~150°C higher than the PCT for the solid fuel at 100% power but 277°C lower than the regulatory limit of 1204°C. When the reflood model is applied to the inner channel, the PCT would be limited to 1100 K (827°C), which is only 50°C higher than the PCT for the solid fuel at 100% power and 377°C lower than the regulatory limit of 1204°C. The calculated fuel temperatures and enthalpies during the REA have been found to be much smaller for the annular fuel, even at 150% power, compared to that for the solid fuel at 100% power. These analyses indicate that the new internally and externally cooled annular fuel can accommodate 50% power uprate in a PWR and still maintain adequate safety margins for a variety of transients and accidents including LOFA, MSLB, LBLOCA, and REA.

CORE-WIDE (IN-PHASE) STABILITY OF SUPERCRITICAL WATER-COOLED REACTORS-II. COMPARISON WITH BOILING WATER REACTORS

To compare the stability features of a supercritical water-cooled reactor (SCWR) design with that of a typical boiling water reactor (BWR), a stability analysis model for a typical BWR has been developed in addition to an already-developed model for the SCWR as presented in a companion paper. The homogenous equilibrium two-phase flow model, which is adequate at high pressures, is applied to the BWR stability analysis. The reactor core is simulated by three channels according to the radial power distribution and the inlet orifice coefficients. Similar to the SCWR model, the neutronic kinetic equation is expanded based on λ modes (reactivity modes). The model is evaluated based on the Peach Bottom Atomic Power Station stability test data, and the results agree well with the experiment. The SCWR is found to be less sensitive to the coolant density neutronic reactivity coefficient than the typical BWR, since most of the neutronic moderation function is provided by the water rods, where the density variation is either zero (if the water rods are insulated) or small (if the water rods are not insulated). The BWR is found to be less sensitive to changes in power level than the SCWR but has the same sensitivity level to the flow rate as the SCWR. A stability envelope that combines the single-channel and in-phase stability modes is developed. The decay ratios for the SCWR together with those for the typical BWR and the new Economic Simplified Boiling Water Reactor at nominal operational conditions are shown in the map. The stability sensitivity to operating conditions is also shown in the map, by increasing the power to 120% of nominal value and decreasing the flow rate to 80% of nominal value. It is found that the SCWR is more sensitive to the single-channel stability compared to the core-wide in-phase stability for all cases.

Coupled Neutronic and Thermal-Hydraulic Out-of-Phase Stability of Supercritical Water Cooled Reactors

Abstract As the last topic of a series of U.S. reference supercritical water-cooled reactor (SCWR) design stability studies, coupled neutronic-thermal-hydraulic out-of-phase stability is analyzed and compared with that of a typical boiling water reactor (BWR). A modal expansion method based on λ modes (reactivity modes) of the neutron kinetic equation is applied, and the first subcritical mode of the neutron dynamics model is coupled with the coolant thermal-hydraulic model. The out-of-phase oscillation of the SCWR is found to be dominated by the reactor thermal hydraulics, whereas the BWR is more sensitive to the coolant density reactivity coefficient because of much stronger neutronic coupling. It is also found that the SCWR stability is sensitive to the details of the core simulation model and the hottest channel dominates the stability. The BWR is less sensitive to the core simulation model since it has much stronger neutronic coupling that is controlled by the whole-core average properties. Power and flow rate sensitivity analysis of the out-of-phase stability was also conducted for both the SCWR and the BWR. The SCWR stability is found to be more sensitive to the operating parameters than the typical BWR. Although the water rod heating can improve the SCWR out-of-phase stability, it cannot significantly improve the sensitivity feature.

Analysis of a Convection Loop for GFR Post-LOCA Decay Heat Removal

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA GFR. The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outperforms helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops.Copyright

Studies of the Deteriorated Turbulent Heat Transfer Regime for the Gas-Cooled Fast Reactor Decay Heat Removal System

Increased reliance on passive emergency cooling using natural circulation of gas at elevated pressure is one of the major goals for the Gas-cooled Fast Reactor (GFR). Since GFR cores have high power density and low thermal inertia, the decay heat removal (DHR) in depressurization accidents is a key challenge. Furthermore, due to its high surface heat flux and low velocities under natural circulation in any post-LOCA scenario, three effects impair the capability of turbulent gas flow to remove heat from the GFR core, namely: (1) Acceleration effect (2) Buoyancy effect (3) Properties variation. This paper reviews previous work on heat transfer mechanisms and flow characteristics of the Deteriorated Turbulent Heat Transfer (DTHT) regime. It is shown that the GFR’s DHR system has a potential for operating in the DTHT regime by performing a simple analysis. A description of the MIT/INL experimental facility designed and built to investigate the DTHT regime is provided together with the first test results. The first runs were performed in the forced convection regime to verify facility operation against well-established forced convection correlations. The results of the three runs at Reynolds numbers 6700, 8000 and 12800 showed good agreement with the Gnielinsky correlation [4], which is considered the best available heat transfer correlation in the forced convection regime and is valid for a large range of Reynolds and Prandtl numbers. However, even in the forced convection regime, the effect of heat transfer properties variation of the fluid was found to be still significant.Copyright

Capability Enhancement of MATRIX Electromagnetic Pump Analysis Code by Including Thermal Analysis

Abstract The legacy electromagnetic (EM) pump analysis tool MATRIX has been improved by the addition of a thermal analysis module. Although the module is patterned after the general-purpose Advanced General Electric Network Analyzer (AGENA) code, it is developed from a more fundamental approach to provide a better understanding and control of the thermal analysis of the EM pump. The MATRIX results are verified against the AGENA results and the test data from the 160 m3/min large EM pump tests, which provided a good estimate of the thermal conductance between the lamination and the inner duct wall. Full and good contact between the lamination and the inner duct wall is necessary to keep the copper conductor temperatures low. Parametric studies, as expected, confirmed the correct trend of increasing copper conductor temperatures with increasing frequency. The MATRIX results show that a new proposed insulation material for the future EM pumps is beneficial since it could reduce the copper block temperature by ~20°C. Such analysis can help develop a better EM pump with a more compact design and better insulation material.

Next Generation Electromagnetic Pump Analysis Tools (PLM DOC-0005-2188). Final Report

This report provides the broad historical review of EM Pump development and details of MATRIX development under this project. This report summarizes the efforts made to modernize the legacy performance models used in previous EM Pump designs and the improvements made to the analysis tools. This report provides information on Tasks 1, 3, and 4 of the entire project. The research for Task 4 builds upon Task 1: Update EM Pump Databank and Task 3: Modernize the Existing EM Pump Analysis Model, which are summarized within this report. Where research for Task 2: Insulation Materials Development and Evaluation identified parameters applicable to the analysis model with Task 4, the analysis code was updated, and analyses were made for additional materials. The important design variables for the manufacture and operation of an EM Pump that the model improvement can evaluate are: space constraints; voltage capability of insulation system; maximum flux density through iron; flow rate and outlet pressure; efficiency and manufacturability. The development of the next-generation EM Pump analysis tools during this two-year program provides information in three broad areas: Status of analysis model development; Improvements made to older simulations; and Comparison to experimental data.