Brian G. Woods

Steady-State Thermal-Hydraulic Analysis of the Oregon State University TRIGA Reactor Using RELAP5-3D

Abstract Oregon State University has recently conducted a complete core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors Program. The goals of the thermal-hydraulic steady-state analysis were to calculate natural-circulation flow rates, coolant temperatures, and fuel temperatures as a function of core power, as well as peak values of fuel temperature, cladding temperature, surface heat flux, critical heat flux ratio, and temperature profiles in the hot channel for both the highly enriched uranium and low-enriched uranium cores. RELAP5-3D Version 2.4.2 was used for all computational modeling during the thermal-hydraulic analysis. This is a lumped parameter code forcing engineering assumptions to be made during the analysis. A single-hot-channel model’s results are compared to results produced from more refined two- and eight-channel models in order to identify variations in thermal-hydraulic characteristics as a function of spatial refinement.

Analyses of the OSU-MASLWR Experimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

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.

Neutronic Analysis of the Oregon State TRIGA Reactor in Support of Conversion from HEU Fuel to LEU Fuel

Abstract In support of the conversion of the Oregon State TRIGA Reactor (OSTR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel, a comprehensive neutronic analysis utilizing MCNP5 was performed on the HEU and LEU core configurations. The initial 1974 HEU core provided an opportunity for verification of the MCNP5 baseline model; all fuel elements in the initial core were congruent in geometry and material composition, having no burnup. In addition, a substantial database of core parameters was documented during the initial HEU core start-up. This verification study examined control rod worth, core excess reactivity, burnup, core power, power per element, temperature coefficient of reactivity, void coefficient of reactivity, moderator coefficient of reactivity, axial and radial power profiles, prompt-neutron lifetime, effective delayed-neutron fraction, power defect, and xenon poisoning. Fuel material composition and core loadings are presented. The excellent comparison between the numerical results and the experimental data of the initial HEU core established an objective, credible baseline model and methodology, which were then extended to the LEU core neutronic analysis. Comparison between the numerically calculated core physics values for the new LEU core and data collected during start-up provided a complete verification that the MCNP5 models developed for both the HEU and LEU cores were representative of the OSTR.

TRACE Code Analyses for the IAEA ICSP on “Integral PWR Design Natural Circulation Flow Stability and Thermo-Hydraulic Coupling of Containment and Primary System During Accidents”

Considering the world energy demand increase in order to fulfill an environmental and economic sustainability, the energy policy of each country has to diversify the sources of energy and use stable, safe energy production option able of producing electricity in a clean way contributing in cutting the CO2 emission. In the framework of the sustainable development, today the use of advanced nuclear power plant, have an important role in the environmental and economic sustainability of country energy strategy. In the last 20 years, in fact, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs considering also the use of natural circulation for the cooling of the core in normal and transient conditions. In this framework, Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a system level test facility to examine natural circulation phenomena characterizing the Multi-Application Small Light Water Reactor (MASLWR) design, a small modular integral pressurized light water reactor relying on natural circulation during both steady state and transient operation. It includes an integrated helical coil steam generator as well. Starting from an experimental campaign in support of the MASLWR concept design verification, the planned work, will be not only to specifically investigate the concept design further but also advance the broad understanding of integral natural circulation reactor plants and accompanying passive safety features as well. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System During Accidents” is hosting at OSU and the experimental data will be developed at the OSU-MASLWR facility. The purpose of this IAEA ICSP is to provide experimental data on single/two-phase flow instability phenomena under natural circulation conditions and coupled containment/reactor vessel behavior in integral-type light water reactors. These data can be used to assess thermal hydraulic codes for reactor system design and analysis as well. The first planned test investigates a stepwise reduction in the primary mass inventory of the facility while operating at reduced power (decay power). The second planned test, investigates a loss of feed water transient with subsequent primary blowdown due to automatic depressurization system actuation and long term cooling phase. The target of this paper is to contribute to the thermal hydraulic analysis of the expected phenomena of these transients on the basis of the TRACE V5 Patch 01 calculated data developed during the double-blind phase of the ICSP.Copyright

A Comparison of Pulsing Characteristics of the Oregon State University TRIGA Reactor with FLIP and LEU Fuel

Abstract The Oregon State TRIGA Reactor (OSTR) was converted from highly enriched uranium (HEU) Fuel Life Improvement Program (FLIP) fuel to low-enriched uranium (LEU) fuel in October 2008. This effort was driven by the U.S. Department of Energys Reduced Enrichment for Research and Test Reactor program. The new LEU fuel is 30/20 U-Zr1.6H (30% uranium in the fuel matrix, 19.75 wt% enriched) in contrast to the FLIP fuel having U-Zr1.6H (8.5% uranium in the fuel matrix, 70 wt% enriched). This new fuel composition provides the best match in performance of the available mixture ratios when compared to the previous FLIP fuel. To support conversion, a complete assessment and reevaluation of the OSTR Safety Analysis Report was performed. This evaluation included steady-state thermal-hydraulic and neutronics characterizations of the HEU and LEU cores as well as a transient behavior (pulse) analysis of both core types. This paper presents a summary of the methods used and results produced during the pulse analysis identifying power, temperature, and reactivity during pulsed operation for the FLIP and new LEU fuel. This analysis was performed using RELAP5-3D version 2.4.2 and point reactor kinetics simulation software; these two methods are found to agree very well. We discuss the differences between the two fuels and the impact of pulse behavior on the safety limits for the converted reactor.

Thermal Analysis and Testing of a Small Radioisotope Power System Concept

Oregon State University (OSU) is conducting an experimental study into the thermal behavior of a GPHS based small RPS concept. The subject RPS configuration is applicable for a number of Mars surface missions, such as using them to power small rovers and small static landers. Each module will use a single GPHS module to generate about 20–25W of electric power. Initial designs for similar RPS concepts have been completed and initial numerical analysis models have been developed by NASA’s JPL. The primary purpose of this research project is to develop an experimental model of the GPHS module based small RPS concept and generate operational data that can be used to validate the thermal analysis codes and methodologies. The validation of codes and methodologies is to be completed by JPL. Five mission phases have been identified for the subject RPS concept. This experimental program focuses on one of these mission phases, earth storage. This paper addresses model design, construction, and testing.

Development and Application of Molecular Tagging Velocimetry for Gas Flows in Thermal Hydraulics

Abstract Molecular tagging velocimetry (MTV) is a nonintrusive velocimetry technique based on laser spectroscopy. It is particularly effective in challenging gas flow conditions encountered in thermal hydraulics where particle-based methods such as particle image (or tracking) velocimetry do not perform well. The main principles for designing and operating this diagnostic are presented as well as a set of gases that have been identified as potential seeds. Two gases [H2O and nitrous oxide (N2O)] have been characterized extensively for thermodynamic conditions ranging from standard temperature and pressure to environments encountered in integral effects test (IET) facilities for high-temperature gas reactors. A flexible, modular, and transportable laser system has been designed and demonstrated with H2O and N2O seed gases. The laser system enables determining the optimum excitation wavelength, tracer concentration, and timing parameters. Velocity precision and thermodynamic domain of applicability are discussed for both tracers. The spectroscopic nature of the diagnostics enables one to perform first-principle uncertainty analysis, which makes it attractive for validating numerical models. Molecular tagging velocimetry is demonstrated for two flows. First, in blowdown tests with H2O seed, the unique laser system enables one of the largest dynamic ranges reported to date for velocimetry: 5000:1 (74 dB). N2O-MTV is then deployed in situ in an IET facility, i.e., the High-Temperature Test Facility at Oregon State University, during a depressurized conduction cooldown (DCC) event. Data enable researchers to gain insights into flow instabilities present during DCC. Thus, MTV shows a strong potential to gain a fundamental understanding of gas flows in nuclear thermal hydraulics and to provide validation data for numerical solvers.

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

The objective of this proposed work is to expand the utilization of the Oregon State University High Temperature Test Facility (HTTF) to the validation of design and safety thermal-hydraulics methods developed for a broader range of advanced high temperature reactors and events. While HTTF was built for a prismatic block High Temperature Gas Reactor (HTGR), it can be expanded to other advanced gascooled reactors as well as key safety aspects of other high temperature reactors such as molten salt and even sodium cooled reactors. The HTTF is based on a 1⁄4-scale model of an HTGR with the capability for both forced and natural circulation flow through the prismatic core with an electrical heat source. The peak core region temperature capability is 1600 ̊C. The facility can be modified to simulate safety-related passive natural circulation core cooling with heat rejection to air as well as passive containment cooling. Both these features are essential to several advanced high temperature reactor concepts including advanced gascooled fast reactors (e.g. EM), molten salt reactors (e.g. AHTR and FHR) and sodium cooled reactors (e.g. PRISM). Although the fluids may be different, the computation methods for flow distribution, coupled natural circulation loops and direct heat rejection to air are similar. This later aspect is an important safety feature of advanced high temperature reactors in that their safety cooling systems reject heat to air through a heat exchanger rather than by evaporative cooling which requires water replenishment within 72 hrs. (e.g. AP1000). Figure 1 shows a schematic of a generic passive natural circulation core cooling system as applied to the current test facility. In addition, there are several advanced gas reactor designs that are radically different in both geometry and operation from the prismatic block HTGR. Two such examples include the pebble bed core type gas reactor and the General Atomics Energy Multiplier Module (EM) reactor. There were several accident scenarios identified in the DOE and NRC gas reactor Phenomenon Identification and Ranking Tables (PIRT) that were not addressed in the original scaling analysis for this test facility. These include reactivity-induced transients, steam-water ingress events and process plant coupling events. It is proposed under this work that the scaling analysis of the HTTF be revised to add a detailed scaling analysis of (1) a passive natural circulation core cooling system, (2) the two advanced gas reactor concepts mentioned above as well as (3) the utilization of the test facility for transients that were not included in the original scaling analysis. It is not expected that the test facility will be able to provide high-quality data for these designs and transients as currently configured and thus this work will also include the development of a set of design requirements, required modifications and a feasibility study for each of these different scaling sets.

Analysis of Primary/Containment Coupling Phenomena Characterizing the MASLWR Design During a SBLOCA Scenario

Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the “Power Reactor Information System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction (IAEA PRIS, 2011).