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

2. 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 TRTR Conference September 19

3. 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 TRTR Conference September 19

4. 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)
[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 TRTR Conference September 19

5. 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 TRTR Conference September 19

6. 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 TRTR Conference September 19

7. 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 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 TRTR Conference September 19

8. 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 TRTR Conference September 19

9. 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 TRTR Conference September 19

10. 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 TRTR Conference September 19

11. 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 TRTR Conference September 19

12. 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 TRTR Conference September 19

13. 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 TRTR Conference September 19

14. 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 kW (HEU-BOL Normal Core)
DNBR
The figure represents the axial DNBR when the hot channel is at 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
Wade Marcum TRTR Conference September 19

15. 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 TRTR Conference September 19

16. 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 TRTR Conference September 19

17. 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 mm
37.34 mm
0.508 mm
673.1 mm
Wade Marcum TRTR Conference September 19

18. 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 TRTR Conference September 19

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.

20. 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

21. 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

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

23. 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.

24. 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
[2] Groeneveld, D.C., S.C.Cheng, and T. Doan, 1986 AECL-UO Critical Heat Flux Lookup Table. Heat Transfer Engineering, (1-2): p
[3] Groeneveld, D.C., et al., The 1995 look-up tables for critical heat flux in tubes. Nuclear Engineering and Design, (23): p
[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.

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