Jose N. Reyes

NuScale Plant Safety in Response to Extreme Events

Abstract The extreme events that led to the prolonged electrical power outage and finally to sever damage of four units of the Fukushima nuclear plant have highlighted the importance of ensuring a technical means for stable, long-term cooling of the nuclear fuel and the containment following a complete station blackout. This paper presents an overview of the advanced passive safety systems designed for the NuScale nuclear power plant and their role in addressing extreme events. The NuScale plant may include up to 12 power modules, and each module incorporates a reactor pressure vessel (core, steam generator, and pressurizer) and a containment vessel that surrounds the reactor vessel. During normal operation, each containment vessel is fully immersed in a water-filled, stainless steel-lined concrete pool that resides underground. The pool, housed in a Seismic Category I building, is large enough to provide 30 days of core and containment cooling without adding water. After 30 days, the core decay heat generation is so small that the natural convection heat transfer to air at the outside surface of the containment, coupled with thermal radiation heat transfer, are completely sufficient to remove the core decay heat for an unlimited period. These passive safety systems can perform their function without requiring an external supply of water or electric power. Computational and experimental assessments of the NuScale passive safety systems are being performed at several institutions, including the one-third scale NuScale integral system test facility at Oregon State University.

The use of MRI to quantify multi-phase flow patterns and transitions : an application to horizontal slug flow

Abstract In this paper we discuss the current benefits and limitations of using magnetic resonance imaging (MRI) to examine multi-phase fluid flow patterns and transitions. The advantages and disadvantages are highlighted in the context of an ongoing collaborative research effort between the University of California, San Francisco (UCSF) School of Medicine MRI Center and the Department of Nuclear Engineering at Oregon State University (OSU). Of particular interest, are the MRI measurements of the liquid volumetric flux distribution, the void fraction and the interfacial area concentration for slug flow in a horizontal air–water system. The data presented herein is unique relative to the slug flow conditions examined and the method implemented for data acquisition. The special scanning sequences designed by UCSF were capable of imaging at repetition intervals as fast as 7 ms.

Anomalous heat output from Pd cathodes without detectable nuclear products

A series of experiments has been conducted to explore the effects of electrolyzing heavy water (D2O) using palladium and platinum electrodes. Over 40 weeks of experimental runs have been conducted in four cells which electrolyze heavy water using palladium and platinum electrodes. Tritium production, neutron and gamma radiation, and cell temperatures were monitored simultaneously and continuously throughout the runs. These experiments have resulted in seven elevated temperature events similar to those claimed by Pons and Fleischmann, with no correlating detection of nuclear products. The seven events which have occurred to date all take the same general form in which the apparent heat output of a cell, as seen in terms of the change in cell fluid temperature, increases in a distinct and significant step. A single light water cells, identical in all respects to those using heavy water, has been operated for over 15 weeks and has produced no temperature excursions, and also no nuclear products.

Testing of Passive Safety System Performance for Higher Power Advanced Reactors

This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

PTS thermal hydraulic testing in the OSU APEX facility

Abstract The APEX Test Facility at Oregon State University has been modified to simulate a typical 2×4 loop Combustion Engineering nuclear plant. The new configuration, APEX-CE, will be used to perform a series of separate effects and integral systems overcooling tests that examine the conditions that lead to primary loop stagnation and cold leg thermal stratification. RELAP5 calculations will be compared to the test data to assess its ability to predict the onset of loop stagnation. A computational fluid dynamics (CFD) code will be assessed against the APEX-CE cold leg and downcomer fluid mixing data. This work is part of the US nuclear regulatory commissions (USNRC) effort to review its existing pressurized thermal shock (PTS) guidance.

The NuScale Advanced Passive Safety Design

NuScale Power is designing an advanced passive nuclear power plant that does not rely on any external sources of power or coolant for safety. Accordingly, the NuScale design inherently prevents the types of issues which led to fuel damage at the Fukushima Daiichi facility. This paper presents an overview of the advanced passive safety systems implemented in the NuScale nuclear power plant. During normal operation, each NuScale containment is fully immersed in a water-filled stainless steel lined concrete pool that resides underground. The pool, housed in a Seismic Category I building, is large enough to provide 30 days of core and containment cooling without adding additional water. After 30 days, the decay heat generation is sufficiently small that natural convection heat transfer to air on the outside surface of the containment coupled with thermal radiation heat transfer is completely adequate to remove core decay heat for an indefinite period of time. These passive safety systems can perform their function without requiring an external supply of water, power, or generators.Copyright

Flow Stagnation under Single and Two-Phase Natural Circulation Conditions in the APEX-CE Test Facility

Natural circulation experiments were conducted at Oregon State University using the advanced plant experiment (APEX-CE) integral system test facility as configured to simulate a typical Combustion Engineering nuclear steam supply system. This paper describes the mechanisms by which natural circulation flow was interrupted under single-phase and two-phase natural circulation conditions in APEX-CE.

ICONE15-10026 Status of the INL Gas Reactor Test System Experiment Facility

The Gas Reactor Test System (GRTS) is an experiment facility for examining the thermal hydraulic performance of the Generation IV, Very High Temperature Reactor (VHTR) during a LargeBreak Loss of Coolant Accident (LB-LOCA). The LB-LOCA is defined as the double guillotine break of the VHTR coaxial inlet and outlet cross duct. Two system safety codes, MELCOR and RELAP5-3D were used to calculate core temperatures and flow rates during the LB-LOCA transient. Computational fluid dynamics modeling of the transient produced flow vectors and gas species distribution. The most important phenomenon during the transient is the lock-exchange process, which suppresses the onset of natural circulation until considerable molecular diffusion has occurred. The GRTS was designed based upon a hierarchical two tier scaling analysis whose primary objective was replicating the lockexchange and natural circulation characteristics of the VHTR. The GRTS uses a scaled graphite core to represent the VHTRs graphite core. An in-depth scaling analysis was performed for the GRTS in order to ensure that it accurately simulated the VHTR thermal responses. RELAP5-3D thermal analyses, ProEngineer stress analyses, and combined FLUENT – STARCD CFD analyses have provided a system design that fulfills the GRTS mission statement. This paper discusses the design analyses and their implications on the GRTS capabilities. A discussion is also presented on the preliminary instrumentation plan. The GRTS will provide an extensive temperature map of the VHTR core outlet plenum and its core support, oxygen transport rates during the lock-exchange phenomenon, and thermal conduction rates from the core to the vessel. As a result of the GRTS using helium coolant at 950 C, the resulting experiment data is expected to considerably extend the U.S. database for hightemperature gas reactor operations. Finally, the discussion will present conclusions from the GRTS manufacturing and quality control processes that may benefit the VHTR design.

Comparative Study of Station Blackout Counterpart Tests in APEX and ROSA/AP600

A comparison is presented between station blackout tests conducted in both the Advanced Plant Experiment (APEX) facility and in the modified Rig of Safety Assessment (ROSA/AP600) Large-Scale Test Facility. The comparison includes the depressurization and liquid-level behavior during secondary-side blowdown, natural circulation, automatic depressurization system operation, and in-containment refueling water storage tank injection. Reasonable agreement between the test results from APEX NRC-2 and ROSA/AP600 AP-BO-01 has been observed with respect to the timing of depressurization and liquid draining rates. This indicates that the reduced height and pressure scaling of APEX preserves the sequence of events relative to the full-height and pressure ROSA/AP600.

A general theory for flooding implementing cuspoids catastrophe

Abstract This paper combines Kelvin-Helmholtz Theory and Catastrophe Theory to develop a general mathematical framework for the flooding phenomenon. The theoretical model proposed in this paper is based on the functional relationship between the gas and liquid flow rates expressed in terms of a modified Kutateladze number that takes into account the effects of entrainment and geometry. A large number of experimental data has been examined against the theoretical model prediction. In most cases, good agreement is obtained by empirically varying only one coefficient.