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

THERMAL HYDRAULIC ANALYSIS FOR THE OREGON STATE

REACTOR USING RELAP5-3D

Wade R. MarcumBrian G. Woods

2007 TRTR ConferenceSeptember 19, 2007

Wade Marcum 2007 TRTR Conference September 19

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

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).



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


Oregon State TRIGA® Reactor Overview

• 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

Lower Grid Plate

Upper Grid Plate

Fuel Elements

Control Element

Oregon State TRIGA® Reactor Isometric and Sectional Isometric Rendering




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 1

Aluminum Clad Reflector Elements 21

Stainless Steel Clad Reflector Elements ---

R

S

R

S

R

S

CT

B-1

B-2B-3

B-4

B-5B-6

C-1

C-2

C-3C-4

C-5

C-6

C-7

C-8

C-9C-10

C-11

C-12

SHIM

REG

TRANS

SAFED-1

D-2

D-3

D-4

D-5D-6D-7

D-8

D-9

D-10

D-11

D-12

D-13

D-14 D-15D-16

D-17

D-18

E-1

E-2

E-3

E-4

E-5

E-6E-7E-8

E-9

E-10

E-11

E-12

E-13

E-14

E-15

E-16

E-17

E-18 E-19 E-20

E-21

E-22

E-23

E-24

F-1

F-2

F-3

F-4

F-5

F-6

F-7F-8F-9

F-10

F-11

F-12

F-13

F-14

F-15

F-16

F-17

F-18

F-19

F-20

F-21

F-22F-23 F-24

F-25

F-26

F-27

F-28

F-29

F-30

G-1

G-2

G-3

G-4

G-5

G-6

G-7

G-8G-9G-10G-11

G-12

G-13

G-14

G-15

G-16

G-17

G-18

G-19

G-20

G-21

G-22

G-23

G-24

G-25

G-26G-27 G-28

G-29G-30

G-31

G-32

G-33

G-34

G-35

G-36

Fuel Element

Reflector Element

FFCR

AFCR

Central Thimble

RabbitTerminus

Source

A-1

CT

HEU FLIP Normal Core Configuration


R

S

R

S

R

S

CT

B-1

B-2B-3

B-4

B-5B-6

C-1

C-2

C-3C-4

C-5

C-6

C-7

C-8

C-9C-10

C-11

C-12

SHIM

REG

TRANS

SAFED-1

D-2

D-3

D-4

D-5D-6D-7

D-8

D-9

D-10

D-11

D-12

D-13

D-14 D-15D-16

D-17

D-18

E-1

E-2

E-3

E-4

E-5

E-6E-7E-8

E-9

E-10

E-11

E-12

E-13

E-14

E-15

E-16

E-17

E-18 E-19 E-20

E-21

E-22

E-23

E-24

F-1

F-2

F-3

F-4

F-5

F-6

F-7F-8F-9

F-10

F-11

F-12

F-13

F-14

F-15

F-16

F-17

F-18

F-19

F-20

F-21

F-22F-23 F-24

F-25

F-26

F-27

F-28

F-29

F-30

G-1

G-2

G-3

G-4

G-5

G-6

G-7

G-8G-9G-10G-11

G-12

G-13

G-14

G-15

G-16

G-17

G-18

G-19

G-20

G-21

G-22

G-23

G-24

G-25

G-26G-27 G-28

G-29G-30

G-31

G-32

G-33

G-34

G-35

G-36

Fuel Element

Reflector Element

FFCR

AFCR

Central Thimble

RabbitTerminus

Source

A-1

CT



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

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




Core Power Distribution

(HEU-BOL Normal Core Configuration)


Core Power Distribution

(HEU-EOL Normal Core Configuration)




0

0.5

1

1.5

2

-20

-10

0

10

200

20

40

60

80

100

120

140

Radial Dist. From Fuel Centerline [cm]Axial Dist. From Fuel Centerline [cm]

Pow

er

[W]

Fuel Rod Power (HEU-BOL Normal Core)

0

0.5

1

1.5

2

-20

-10

0

10

200

20

40

60

80

100

120

Radial Dist. From Fuel Centerline [cm]Axial Dist. From Fuel Centerline [cm]

Pow

er

[W]

Fuel Rod Power (HEU-EOL 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.




These two vectors were then input into the RELAP5-3D model

0

0.5

1

1.5

2

2.5

3

3.5

0 0.5 1 1.5 2

Radial Distance From Fuel Centerline [cm]

Pow

er F

acto

r

HEU-BOL Normal

HEU-EOL Normal

Radial Power Factor Distribution

0

0.2

0.4

0.6

0.8

1

1.2

1.4

-20 -15 -10 -5 0 5 10 15 20

Axial Distance From Fuel Centerline [cm]

Pow

er F

acto

r

HEU-BOL NormalHEU-EOL Normal

Axial Power Factor Distribution


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


HEU Benchmark Results – Steady State (BOL)


400

450

500

550

600

650

700

750

800

0 0.005 0.01 0.015 0.02

Radial Distance From Fuel Element Center [m]

.05 mil Gap

.10 mil Gap

.15 mil Gap

.20 mil GapIFE Measurement

Fuel Element Radial Temperature Distribution at 1.1 MWthT

empe

ratu

re [

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.




0

100

200

300

400

500

600

700

800

900

1000

0 10 20 30 40

Hot Channel Fuel Element Power [kW]

0.02

0.04

0.06

0.08

0.10

0.12

0.14

0.16

Fuel CenterlineOuter CladdingBulk CoolantCoolant Mass Flow

Hot Channel Properties (HEU-BOL Normal Core)

Tem

pera

ture

[K

]

Coo

lant

Mas

s Fl

ow R

ate

[kg/

sec]

Hot Channel MDNBR (HEU-BOL Normal Core)A

xial

MD

NB

R




Hot Channel Axial DNBR at 18.02 kW (HEU-BOL Normal Core)D

NB

R

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.


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].

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

0.000 0.010 0.020 0.030 0.040 0.050

Time [sec]

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

$2.60 Power $2.70 Power $2.90 Power

$2.60 Energy $2.70 Energy $2.90 Energy

Pow

er [

MW

]

Ene

rgy

[MJ]

HEU-EOL Normal Core Pulse Trace




,

0.18 0.009HEU LEU FUEL

K T

3

,2.04 4.17 10P HEU LEU FUEL

C T T

[W-sec/cm3-C]

[W/cm-C]

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:

300

400

500

600

700

800

900

1000

1100

1200

2.50 2.60 2.70 2.80 2.90

Reactivity Insertion [$]

0

2000

4000

6000

8000

10000

12000

14000

Max. Core Avg. Temp. (Fuch's)Max. Hot Rod Temp. (RELAP5-3D)Max. Hot Rod Temp. (Current SAR)Fuel Temp LimitPulse Peak Power

Pulse Results (HEU-EOL Normal Core)

Puls

e Pe

ak P

ower

[M

W]

Tem

pera

ture

[C

]




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


Experimental Measurements


Equipment Brand Model Number Serial Number

Thermocouple Gordon K Type G16308

GPIB/USB Interface Cable HP Agilent 82379A X1307A2

DAQ Bucket HP Agilent 34970A 11298

Computer Dell E1705 F5PF1B1


• 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.



• 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



0

5

10

15

20

25

30

35

40

45

50

0 10 20 30 40 50 60 70

Axial Distance From Lower Grid Plate [cm]

Tem

pera

ture

[°C

]

Experimental

RELAP5-3D

Local Axial Bulk Coolant Temperature Distribution


20

25

30

35

40

45

50

55

60

65

70

0 10 20 30 40 50 60 70

Axial Distance From Lower Grid Plate [cm]

Tem

pera

ture

[°C

]

Local Axial Bulk Coolant Relative Temperature Change

RELAP5-3D (HEU-MOL Normal Core Configuration) was used when comparing the coolant temperature change as a function of axial position.


Local Axial Temperature Distribution

20

25

30

35

40

45

50

55

60

65

70

0 10 20 30 40 50 60 70

Axial Distance From Lower Grid PLate [cm]

Tem

pera

ture

[°C

] Loc. 1

Loc. 2

Loc. 3

Loc. 4

Loc. 5

Loc. 6


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.


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.

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