THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Wade R. Marcum Brian G. Woods 2007 TRTR Conference September 19, 2007

Project Introduction/Objective Benchmark Methodology THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Outline Project Introduction/Objective Benchmark Methodology Oregon State TRIGA® Reactor Overview RELAP5-3D Model HEU Benchmark Results Steady State (BOL) Pulse (EOL) Experimental Measurements Conclusion Wade Marcum 2007 TRTR Conference September 19

Project Introduction/Objective THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Project Introduction/Objective Oregon State University has completed its core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors (RERTR) program. As part of the core conversion analysis, the Nuclear Regulatory Commission (NRC) has required that complete neutronic and thermal hydraulic analyses be conducted on the existing OSTR (HEU) and potential (LEU) core. The goals of the thermal hydraulic analyses were to: Calculate natural circulation flow rates, coolant temperature and fuel temperatures as a function of core power for both the HEU and LEU cores. For steady state and pulsed operation, calculate peak values of the fuel temperature, cladding temperature, surface heat flux as well as departure from nucleate boiling ratio (DNBR) and temperature profiles in the hot channel for both the HEU and LEU cores. Perform accident analyses for the accident scenarios identified in the OSTR Safety Analysis Report (SAR). Wade Marcum 2007 TRTR Conference September 19

Cross Sectional View of Instrumentation Fuel Element THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Benchmark Methodology Steady State Operation (1.1 MWth) Beginning of Life Core Instrumentation Fuel Element (IFE) 356-373 [C] Pulse Operation End of Life Core Reactivity Insertion/Max Fuel Temp. $2.60/1100 [C] $2.70/1150 [C] Effective Peak Factor (Current SAR) 3.41 Cross Sectional View of Instrumentation Fuel Element Zirconium Pin Fuel Gap Stainless Steel Clad 0.762 cm Thermocouple Wade Marcum 2007 TRTR Conference September 19

Oregon State TRIGA® Reactor Overview THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Oregon State TRIGA® Reactor Overview Lower Grid Plate Upper Grid Plate Fuel Elements Control Element Oregon State TRIGA® Reactor Isometric and Sectional Isometric Rendering TRIGA Mark II Reactor In operation since 1976 Circular lattice fuel rod configuration Core located ~5 meters below surface Pool depth ~6 meters Current fuel: HEU Fuel Lifetime Improvement Plant (FLIP) Fuel Wade Marcum 2007 TRTR Conference September 19

Core Components for HEU FLIP Fuel THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Oregon State TRIGA® Reactor Overview HEU FLIP Normal Core Configuration Core Components for HEU FLIP Fuel Core Configuration HEU FLIP Standard Fuel Elements 81 Instrumented Fuel Assemblies 1 Fuel-Followed Control Rod 3 Void-Followed Transient Rod Aluminum Clad Reflector Elements 21 Stainless Steel Clad Reflector Elements --- Wade Marcum 2007 TRTR Conference September 19

Oregon State TRIGA® Reactor Overview THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Oregon State TRIGA® Reactor Overview TRIGA® Fuel Element Design Utilized in the OSTR Core. 87.38 mm 381 mm 88.138 mm 37.34 mm 0.508 mm 673.1 mm Comparison of HEU FLIP Fuel Designs Fuel Type HEU FLIP Uranium content [mass %] 8.5 U-235 enrichment [mass % U] 70 Erbium content [mass %] 1.6 Fuel alloy inner diameter [mm] 6.35 Fuel alloy outer diameter [mm] 36.449 Fuel alloy length [mm] 381 Cladding material Type 304 SS Cladding thickness [mm] 0.508 Cladding outer diameter [mm] 37.465 Wade Marcum 2007 TRTR Conference September 19

Oregon State TRIGA® Reactor Overview THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Oregon State TRIGA® Reactor Overview Core Power Distribution (HEU-BOL Normal Core Configuration) Core Power Distribution (HEU-EOL Normal Core Configuration) Wade Marcum 2007 TRTR Conference September 19

Oregon State TRIGA® Reactor Overview THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Oregon State TRIGA® Reactor Overview Fuel Rod Power (HEU-EOL Normal Core) Fuel Rod Power (HEU-BOL Normal Core) The hot rod fuel element power distribution for the reference core configurations was normalized into two vectors, a radial and axial discontinuous function. Wade Marcum 2007 TRTR Conference September 19

Oregon State TRIGA® Reactor Overview THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Oregon State TRIGA® Reactor Overview Radial Power Factor Distribution Axial Power Factor Distribution These two vectors were then input into the RELAP5-3D model Wade Marcum 2007 TRTR Conference September 19

THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D RELAP5-3D Model Coolant Source Cold Leg Horizontal Connector Hot Channel Coolant Sink Schematic of RELAP5-3D model Geometric/Hydraulic Hot Channel Inputs Implemented in RELAP5-3D Geometric/Hydraulic Description Value Unheated core length at inlet [m] 0.1655 Unheated core length at outlet [m] 0.1647 Inlet pressure loss coefficient 1.29 Exit pressure loss coefficient 0.574 Absolute pressure at the top of the core [Pa] 1.49E5 Inlet coolant temperature [C] 49.0 Flow area [m2] 3.304E-04 Wetted perimeter [m] 0.1177 Hydraulic diameter [m] 2.051E-02 Fuel element heated length [m] 0.381 Fuel element surface area [m2] 3.810E-1 Fuel element surface roughness [m] 2.134E-06 Wade Marcum 2007 TRTR Conference September 19

Fuel Element Radial Temperature Distribution at 1.1 MWth THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D HEU Benchmark Results – Steady State (BOL) Fuel Element Radial Temperature Distribution at 1.1 MWth Temperature [K] A radial temperature profile at 1.1 MWth integral core steady state power was mapped while varying the fuel to clad gap from 0.05 to 0.20 mils, the corresponding temperature was compared to that found in the IFE during the original 1976 core configuration. As a result of this figure a clad gap of 0.1 mils was used in all core configurations. Wade Marcum 2007 TRTR Conference September 19

HEU Benchmark Results – Steady State (BOL) THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D HEU Benchmark Results – Steady State (BOL) Hot Channel Properties (HEU-BOL Normal Core) Temperature [K] Coolant Mass Flow Rate [kg/sec] Hot Channel MDNBR (HEU-BOL Normal Core) Axial MDNBR Wade Marcum 2007 TRTR Conference September 19

Hot Channel Axial DNBR at 18.02 kW (HEU-BOL Normal Core) THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D HEU Benchmark Results – Steady State (BOL) Hot Channel Axial DNBR at 18.02 kW (HEU-BOL Normal Core) DNBR The figure represents the axial DNBR when the hot channel is at 18.02. The methods for calculating DNBR shown use the results produced from RELAP5-3D to apply the appropriate correction factors used in the Groeneveld 1986 [1], 1995 [2], and 2006 [3] AECL-UO look-up tables. The MDNBR value produced from the bounding DNBR method, Groeneveld 2006, is 3.420. Wade Marcum 2007 TRTR Conference September 19

HEU-EOL Normal Core Pulse Trace THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D HEU Benchmark Results – Pulse (EOL) The pulsing performance of the reactor was analysed using a point reactor kinetics model. With this methodology and the fissile fuel characteristics produced from the MCNP5 analysis a pulse power trace was developed for given reactivity insertions [4]. Power [MW] Energy [MJ] HEU-EOL Normal Core Pulse Trace Wade Marcum 2007 TRTR Conference September 19

Pulse Results (HEU-EOL Normal Core) THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D HEU Benchmark Results – Pulse (EOL) Applying power traces respectively, the figure below was produced. During the pulse analysis the fuel thermo-physical properties inherently dominate the tendency for peak fuel temperature over heat removal capability of the system, therefore the thermal conductivity and volumetric heat capacity [5] are as follows: Thermal Conductivity: Volumetric Heat Capacity: Pulse Results (HEU-EOL Normal Core) Pulse Peak Power [MW] Temperature [C] [W/cm-C] [W-sec/cm3-C] Wade Marcum 2007 TRTR Conference September 19

TRIGA® Fuel Element Design Utilized in the OSTR Core. THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D HEU Benchmark Results – Pulse (EOL) TRIGA® Fuel Element Design Utilized in the OSTR Core. 87.38 mm 381 mm 88.138 mm 37.34 mm 0.508 mm 673.1 mm Wade Marcum 2007 TRTR Conference September 19

Experimental Measurements THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Experimental Measurements Equipment Brand Model Number Serial Number Thermocouple Gordon K Type G16308 GPIB/USB Interface Cable HP Agilent 82379A X1307A2 DAQ Bucket 34970A 11298 Computer Dell E1705 F5PF1B1 Wade Marcum 2007 TRTR Conference September 19

Experimental Measurements THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Experimental Measurements This analysis was conducted to provide general coolant temperature profile trends within the OSTR core in order to compare values to the OSTR RELAP5 model. Axial temperature distributions at six different radial locations within the core were produced during this analysis. The OSTR core fuel configuration is skewed symmetrically to one side; it is assumed that this produces hotter coolant temperature values along this radial direction than generally found elsewhere in the core.

Experimental Measurements THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Experimental Measurements A LabVIEW program was developed to collect real-time data samples (sample rate of 2 Hz) In each core radial location (i.e. 1 through 6), 14 axial temperature measurements were taken. Each temperature measurement collected 200 samples over a period of 100 seconds The 14 axial temperature measurements were taken evenly between the lower and upper grid plate by incrementing the measurements every 2 inches

Experimental Measurements THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Experimental Measurements Local Axial Bulk Coolant Temperature Distribution RELAP5-3D (HEU-MOL Normal Core Configuration) was used when comparing the coolant temperature change as a function of axial position. Local Axial Bulk Coolant Relative Temperature Change

Experimental Measurements THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Experimental Measurements

THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Conclusion The steady state fuel temperature at 1.1 MWth reflects that found in the IFE by applying a fuel/clad contact gap thickness of 1.0 mils to within a relative error margin. The hot channel maximum fuel temperature during a pulse reflects that in the current SAR within ~100 [C]. The experimental temperature measurements taken provide evidence that the RELAP5-3D model produces conservative results to that found in the physical OSTR. Through the benchmark methodology presented, the thermal hydraulic analysis conducted during this core conversion projects produces conservative and relatively accurate results.

THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D References [1] Groeneveld, D.C., et al., The 2006 CHF look-up table. Nuclear Engineering and Design, 2007: p. 1-24. [2] Groeneveld, D.C., S.C.Cheng, and T. Doan, 1986 AECL-UO Critical Heat Flux Lookup Table. Heat Transfer Engineering, 1986. 7(1-2): p. 46-62. [3] Groeneveld, D.C., et al., The 1995 look-up tables for critical heat flux in tubes. Nuclear Engineering and Design, 1996. 1(23): p. 1-23. [4] Safety analysis report for the conversion of the Oregon State TRIGA Reactor from HEU to LEU fuel, in Documentation of analyses of conversion of the Oregon State University TRIGA reactor from HEU to LEU fuel. 2007, Oregon State University: Corvallis. [5] Simnad, M., F. Foushee, and G. West, Fuel elements for pulsed TRIGA Research Reactors. 1975, General Atomics: Sandiego, CA.

THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE REACTOR USING RELAP5-3D Questions