Critical heat flux tests for a 12 finned-element assembly

Nuclear Engineering and Design 313 (2017) 129–140

Contents lists available at ScienceDirect

Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes

Critical heat flux tests for a 12 finned-element assembly J. Yang ⇑, D.C. Groeneveld, L.Q. Yuan Thermalhydraulics Branch, Canadian Nuclear Laboratories, Chalk River, Ontario K0J 1J0, Canada

h i g h l i g h t s
CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions.
Test approach to maximize experimental information and minimize heater failures.
Three series of tests were completed in vertical upward light water flow.
Bundle simulators of two axial power profiles and three heated lengths were tested.
Results confirm that the prediction method predicts lower CHF values than measured.

a r t i c l e

i n f o

Article history: Received 29 June 2016 Received in revised form 1 December 2016 Accepted 3 December 2016

Keywords: Critical heat flux NRU reactor driver fuel Test trajectory

a b s t r a c t An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of each test series were performed following a specific heat flux vs. local subcooling test trajectory, which was designed to allow for measurements at sufficiently wide local conditions. The experimental results confirmed that the current CHF prediction method predicts lower CHF values than experimentally measured using the NRU driver fuel rod simulators. This paper introduces the experimental approach, describes the experiment and presents the experimental data and analysis results. Crown Copyright Ó 2016 Published by Elsevier B.V. All rights reserved.

1. Introduction 1.1. Background The National Research Universal (NRU) reactor is a 135 MWt nuclear research reactor located at the Canadian Nuclear Laboratories (CNL). It is a multipurpose science facility that serves three ⇑ Corresponding author. E-mail address: [email protected] (J. Yang). http://dx.doi.org/10.1016/j.nucengdes.2016.12.010 0029-5493/Crown Copyright Ó 2016 Published by Elsevier B.V. All rights reserved.

main roles: to be a supplier of industrial and medical radioisotopes used for the diagnosis and treatment of life-threatening diseases; to be a major Canadian facility for neutron physics research, and to provide engineering research and development support for CANDUÒ1 power reactors. The NRU reactor core consists of four different types of fuel bundles (or assemblies): driver fuel bundle, Moly-99 rod assembly, 1 CANDUÒ (CANada Deuterium Uranium) is a registered trademark of Atomic Energy of Canada Limited (AECL).

130

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

Nomenclature CHF G _ m MMC P q00

critical heat flux, kW/m2 mass flux, kg/m2/s mass flow rate, kg/s CHF as calculated from modified Menegus correlation, kW/m2 pressure, kPa (a) or psia heat flux, kW/m2

Acronyms AFD axial heat flux distribution CHF critical heat flux CRL Chalk River Laboratories DAS data acquisition system DNB departure from nucleate boiling DP differential pressure transmitter

Mk-4 Fast Neutron (FN) rod assembly, and Mk-7 FN rod assembly. Due to its high operating power, the driver fuel bundle is considered to be the most limiting in terms of critical power under the given flow conditions, i.e., the power corresponding to the initial occurrence of critical heat flux (CHF). The NRU driver fuel assembly consists of twelve fuel elements (pins) arranged in two rings – three elements in the inner ring and nine in the outer ring. Each fuel element is 2.92 m long, consisting of a fuel core (heating zone) 2.74 m in length and a 90 mm long aluminium plug at each end. The fuel core contains high-density uranium silicide (U3Si) in a continuous aluminium matrix; it is extrusion clad with a finned aluminium alloy sheath and is hermetically sealed with welded aluminium end plugs. Each element has 6 fins. The twelve elements are supported by a bottom hanger plate. Six flow spacer plates are placed at regular intervals to position the fuel elements over the remainder of their length. The assembly is located inside a 50-mm ID flow tube and cooled with an upward flow of heavy water (D2O). Fig. 1 illustrates the cross section of the assembly. CHF for the NRU reactor driver fuel is currently evaluated using the modified Menegus correlation (MMC). This correlation is based primarily on the Menegus correlation (Menegus, 1959), corrected for the hydraulic diameter effect observed from CHF measurements on finned single-element and finned fuel bundle simulators.

Fig. 1. Cross-section geometry of NRU driver fuel assembly.

FES ID LB LOFA LORA MMC NRU PCD PT RFD RTD

fuel element simulator inside diameter lower bound loss-of-flow accident loss-of-regulation accident modified Menegus correlation National Research Universal pitch circle diameter pressure transmitter or pressure tap radial heat flux distribution resistance temperature device

Subscript exp experimental value nor nominal value

The Menegus correlation was derived with CHF data obtained with bare cylindrical rods in annular channels. Predictions of CHF, have been further reduced to account for uncertainties in measurements of plant parameters and CHF calculations. These reduced predictions are referred to as lower-bound values of the modified Menegus correlation, which have been applied in safety analyses of the NRU driver fuel during normal operation and postulated accidents. However, there is currently no CHF data available for full-scale NRU driver fuel bundle to validate the CHF prediction. An experimental program was hence initiated to provide relevant data for NRU driver fuel assembly in support of continuing operation of the NRU reactor.

1.2. Test objective The objective of the experiment was to validate the CHF prediction for the NRU driver fuel assembly when using the lower-bound of MMC for safety analysis, i.e., to confirm no CHF occurrence below lower-bound values of MMC. To satisfy the objective, the proposed experiment had to be performed using electrically heated simulators simulating the geometry and radial and axial power profiles or heat flux distributions (RFD and AFD) of the drives fuel assembly as closely as possible. The ranges of the experimental flow conditions and heat flux levels had to bound the ranges of interest for the analysis of postulated slow loss-ofregulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. This required tests at very high heat fluxes, very high mass velocities, and high subcoolings where the CHF mechanism is the departure from nucleate boiling (DNB). Generally, the occurrence of CHF is characterized by a sharp reduction of the local heat-transfer coefficient and hence an increase of the surface temperature. However, the surface temperature behaviour at and after CHF occurrence strongly depends on the CHF mechanism which is in turn a function of the flow channel geometry and local flow conditions. For the dryout type of CHF corresponding to high quality at CHF location, the surface temperature rise after CHF is moderate and controllable, and hence the tested bundle simulator can be reused to obtain a large amount of CHF data points covering a wide range of flow conditions; this is the case for the regular CHF tests of the CANDUÒ power reactors.2 For the DNB type of CHF, however, failure of the heating surface 2 In practice, for the CHF tests in water with CANDUÒ bundles, CHF was considered to occur when one or more of the bundle thermocouples indicated that the surface temperature exceeded the steady nucleate boiling surface temperature by about 20 °C.

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

(a) 0.6-m Heated-Length Bundle

131

(b) 0.8-m Heated-Length Bundle

Fig. 2. Photographs of the 0.6-m and 0.8-m heater rods after CHF tests.

due to rupture or melting (physical burnout) is expected when CHF is reached in an experiment with a fuel element simulator (see Fig. 2), which is challenging from an experimental point of view. A difficulty also arises from the possible bundle failure prior to reaching CHF or caused by premature CHF occurrence due to element bowing or flow oscillations. In addition to the technical difficulties in performing the tests, such CHF experiments are very costly and time consuming since a large quantity of bundle simulators would be needed for the tests to cover the entire range of relevant conditions, given that physical burnout would be resulted from the CHF occurrence. 1.3. Experimental approach An experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. To reach the required range of heat fluxes, test bundles of several heated lengths with uniform and non-uniform AFDs were employed: 1-m and 0.6-m for the uniform AFD bundle and 0.8-m for the non-uniform AFD bundle. Using a heated length shorter than that for driver fuel in NRU and the uniform AFD is acceptable since the NRU reactor is operated at highly subcooled conditions where CHF is basically a local phenomenon, unaffected by upstream conditions (Yang et al., 2006). The local conditions approach implies that subcooled CHF prediction methods (i.e. the Menegus equation (Menegus, 1959) or any other good subchannel CHF prediction method) are a function only of the local pressure, mass flux, subcooling and cross sectional geometry and are independent of upstream AFD. For the above reasons, the present test program focuses on the uniform AFD bundle and uses the nonuniform AFD bundle for confirmation of the insignificancy of the AFD effect. Three series of tests on electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. The non-uniform axial power profile employed in the third series of tests represents the critical section (where CHF has been predicted) of the limiting profile (i.e., the AFD leading to the minimum critical power ratio) among various possible axial power profiles during the postulated loss-of-regulation event. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow LORA and LOFA scenarios.

To allow for measurements at a sufficiently wide range of local subcoolings, a specific test trajectory (a diagram showing the test procedure and test conditions, etc.) was developed for each mass flow rate of each test series. During previous tests on finned element bundles at CNL, a limited amount of information on CHF occurrence was obtained, since the power to the bundle was simply increased for a fixed set of inlet conditions until the bundle failed at the occurrence of CHF. As stated before, the main objective of the present test program is to verify that the current lower-bound predictions of CHF for the NRU driver fuel are adequate for safety analysis (i.e., no CHF occurrence below lower-bound values). Accordingly, the test trajectory is designed to allow a plenty of extra experimental information under the predicted limiting conditions to be obtained at various local subcoolings using each bundle simulator, in addition to the CHF point. The conditions corresponding to the lower-bound predictions of CHF using MMC is hereby referred to as limiting conditions. Fig. 3 shows examples of the planned test trajectories for 0.8-m heated-length non-uniform AFD bundle at mass flow rates of 14.7 kg/s and 8.8 kg/s (corresponding to the analysis of postulated slow LORA and LOFA scenarios, respectively). These trajectories are designed on a heat flux vs. local subcooling basis with a number of heat flux levels. Each point of a trajectory represents the local conditions at the downstream end of the bundle simulator heated length. As indicated in the test trajectory, testing was started at the lowest achievable inlet fluid temperature followed by gradually increasing the power to the test section until the first (lowest) heat flux level has been reached. The inlet temperature was then gradually increased until the lower bound MMC CHF value has been reached at this heat flux (the predicted limiting conditions). After recording this limiting data point, the inlet temperature was gradually reduced back to the minimum achievable value, followed by increasing the heat flux to the second heat flux level and increasing the inlet temperature to reach the second limiting conditions. The procedure was repeated until the limiting data point at the last (the maximum achievable) heat flux has been recorded. At the end, the inlet temperature was further increased to approach the MMC curve and CHF condition (i.e., test section melting). Following the test trajectory, a number of limiting data points relevant to the CHF prediction at various local subcoolings can be obtained using each fuel simulator, in addition to the CHF point. 2. Description of the experimental set up CHF tests on NRU driver fuel simulators were performed in the MR-2 loop at CNL. Fig. 4 shows a schematic of the MR-2 flow loop

132

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140 6000

4000

3000

4000

3000

2000

2000

1000

1000

start

start

0 120

AFD Profile LOFA Modified Menegus Correlation Lower Bound MMC Test Trajectory. L = 0.8 m, W = 8.8 kg/s Example of the AFD Profile of the Simulator

5000 Heat Flux, kW·m-2

5000 Heat Flux, kW·m-2

6000

AFD Profile LORA Modified Menegus Correlation Lower Bound MMC Test Trajectory. L = 0.8 m, W = 14.7 kg/s Example of the AFD Profile of the Simulator

100

80

60 40 Local Subcooling, K

20

0

(a) for 0.8-m Heated-Length Bundle at 14.7 kg/s

0 120

100

80

60 40 Local Subcooling, K

20

0

(b) for 0.8-m Heated-Length Bundle at 8.8 kg/s

Fig. 3. Examples of planned test trajectories.

heat from the outlet water, thus allowing the pump to run at a lower temperature. The water used in the primary side of the loop is de-ionized periodically and is filtered when used in the loop. To reduce the possible presence of void at the outlet during the high-heat-flux low-subcooling tests, cold water was injected into the outlet plenum from a test section bypass line. This minimized large pressure drops at the outlet – a frequent source of flow instability at 2-phase outlet conditions. 2.1. NRU driver fuel bundle simulator

Fig. 4. Schematic of MR-2 loop.

used for these tests. The loop consists of a water reservoir, a pump, the test-section, a heat exchanger, various cooling and heating lines, and the associated piping. The loop can accommodate the high NRU flow rate (i.e., 14.7 kg/s) and powers up to 1.7 MW. Net heat rejection is provided by a plate-type heat exchanger; a second plate-type heat exchanger is used to pre-heat the inlet water with

The test section simulates a NRU driver fuel assembly inside the aluminium (50-mm ID) flow tube. The NRU driver fuel assembly consists of twelve fuel elements (pins) arranged in two rings – three in the inner ring with a Pitch Circle Diameter (PCD) of 13.0 mm, and nine in the outer ring with a PCD of 34.4 mm (see Fig. 1). Each fuel element is 2.92 m long, and consists of a 2.74-m long uranium-aluminium fuel core (heating zone) with a 90-mm long aluminium plug at each end. The fuel core is inside extrusion clad with a finned aluminium alloy sheath. Each element is equipped with 6 fins. The twelve elements are tied to a bottom hanger plate. Six grid spacers are placed at regular intervals (476.3 mm) along the assembly. 2.1.1. Overall geometry of the bundle simulator The tested bundle simulated a NRU driver fuel bundle, except for its heated length. Each bundle simulator consisted of twelve

Flow tube

Spacer

(a) NRU Grid Spacer (b) Low Impact Grid Spacer inside the Flow Tube Fig. 5. Schematic of grid spacers.

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

Fig. 6. Test station setup and instrumentation locations.

133

134

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

NRU fuel element simulators (FESs) and was located inside the NRU aluminium flow tube (50-mm ID). The configuration of the fuel bundle simulator was identical to the NRU driver fuel bundle as shown in Fig. 1. To maintain the correct element spacing, actual NRU grid spacers were used (each spacer occupies 29.3% of the undisturbed cross-sectional flow area, see Fig. 5a) and placed at regular intervals (476.3 mm). Joule heating was introduced to simulate the power generation. Because of large currents (maximum current: about 1000 A per element), large magnetic forces (estimated at 80 N/m of element length) were induced, potentially deforming the electrically heated elements towards the centre. This change in simulator geometry was non-representative of the fuel assembly in the NRU reactor and was an artefact of joule heating. Therefore, additional spacers were introduced midway between adjacent NRU spacer locations to prevent the bundle simulator from deforming and to maintain the assembly geometry. Each of these additional grid spacers occupied 11.5% of the undisturbed cross-sectional flow area (see Fig. 5b, as compared to 29.3% for the NRU spacer) and had rounded leading and trailing edges. They presented a minimum flow disturbance to the coolant and hence had little (if not negligible) impact on CHF at possible CHF locations (i.e., the end of the heated length) (see Section 5.4 for detailed discussion). Fig. 6 shows the test station setup and instrumentation locations. 2.1.2. Geometry of the fuel element simulators An electrical-resistance-heated Inconel 718 tube was used to simulate the power generation of the fuel core in a NRU driver fuel element. Fig. 7 shows the cross section of the fuel element simulator (i.e., the heater rod). Each heater rod consisted of a heater tube, a copper extension rod at the inlet and outlet ends brazed to the heater tube, a plasma-sprayed alumina coating on the outer surface of the heater tube and copper extension rods, and the NRU finned aluminium sheath extruded over the assembly of heater rod and copper extension rods. The alumina coating provides electrical isolation between heater tube and cladding. The NRU finned aluminium sheath was extruded over the Inconel heater tube at the Nuclear Fuel Fabrication Facility of CRL, which manufactures

the actual NRU driver fuel rods, using the same manufacturing process. Heater rods with 1.0 m, 0.8 m and 0.6-m heated lengths were manufactured for the experiment. Fig. 8 shows the overall dimensions of the heater rod having a heated length of 0.8 m. To eliminate the hydrodynamic entrance effect, an unheated length of respectively 0.2 m, 0.4 m, and 0.6 m was located upstream of the 1.0 m, 0.8 m, and 0.6 m heated-length test sections. All heater rods, irrespective of their heated length, have the same external dimensions. 2.1.3. Power profiles of the bundle simulator All test bundles simulated the radial power profile corresponding to the fresh fuel, which has been shown to be the most limiting in term of power at which CHF is observed. The ratio of element power to bundle-average element power is 1.031 for outer-ring elements and 0.906 for inner-ring elements. The CHF mechanism corresponds to departure from nucleate boiling (DNB) for the NRU driver fuel. It is independent of the axial power profile (i.e., local heat flux is the dominant factor) (Yang et al., 2006). The axial and radial power profiles were achieved by changing the heater tube wall thickness (changing the tube inside diameter at the fixed outside diameter) and hence the electrical resistance along the heated length. The heater impedances closely matched that of the MR-2 loop power supply of the Thermalhydraulics Lab at CRL. The maximum achievable power is 1.7 MW with closely matched electrical resistance of the bundle to either the current limit or the voltage limit of the power supply. Three heated lengths (1-m, 0.8-m and 0.6 m) were used for the heaters to obtain data over the test range of operating conditions. Bundle simulators having the 1.0-m and 0.6-m heated lengths were designed with uniform axial power profile to expedite the experiment (manufacturing of FESs with a non-uniform axial power profile is more complex and time-consuming). A separate simulator having a non-uniform axial power profile was constructed to provide confirmatory data for the CHF prediction. It simulated the critical section of the limiting axial power profile, where CHF occurrence was predicted in the safety analysis, and

Ø0.762±0.08

0.030 ±0.003

CLADDINGTHICKNESS THICKNESS Ø0.76±0.08 CLADDING 0.030 +/- 0.003 0.050 ±0.003 Ø1.27±0.08

HEATER ID (vary) HEATER OD Ø5.33

ALUMINA COATING THICKNESS 0.08

Ø5.49REF. REF Ø0.216 Ø7.00±0.05

Ø0.276±0.002

(a) NRU FUEL-PIN CLAD

(b) HEATER TUBE

Note: Dimensions are in inches.

Heated Length (m) 1 0.6 0.8

Outer-Ring Heater ID (mm) 4.22 4.69 0.41 – 0.53

Inner-Ring Heater ID (mm) 4.37 4.78 0.36 – 0.46

Fig. 7. Cross section of NRU driver fuel element simulator.

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

135

Temperatures: Loop and test-section temperatures were measured using K-Type thermocouples and resistance temperature devices (RTDs). Loop Flow: The mass flow rate was measured using a new Cameron/Barton 7203 turbine flow meter with a built-in MC-III WP flow analyzer (transmitter). Pressure Transmitters: Absolute pressures at the test section inlet and outlet and differential pressures (DPs) along the test section were measured. Pressure-tap locations are specified in Fig. 6. Power: The power was obtained from the measured voltage drop across, and the current through, each FES (see Fig. 10). Voltage taps were installed on copper extension rods at both ends of the FES providing a direct power path to each element. Twelve appropriately ranged current shunts with a standard accuracy of ±0.25% (as specified by the manufacturer), one for each FES, were installed in the power lines. The following measurements were recorded during the bundle tests:











Total power, Current through each heater (12 heaters in total), Current through main shunt, Voltage drop, Inlet pressure, Outlet pressure, Pressure drops (five DP cells), Inlet temperature (duplicated measurements), Outlet temperature (duplicated measurements), and Flow rate.

The uncertainties for the measured and derived parameters were determined. A summary of these uncertainties is presented in Table 1. Fig. 8. Heater rod with an 800-mm heated length (dimensions in mm).

3. Test conditions

Local-to-Average Heat Flux Ratio (-)

1.4 1.2 1 0.8 Full Length Portion to Be Tested

0.6 0.4 0.2 0 0

0.5

1 1.5 Axial Location (m)

2

2.5

Fig. 9. Non-uniform axial power profile for the 0.8 m heater superimposed on the full length NRU fuel assembly power profile.

has an overall heated length of 0.8 m. The non-uniform axial power profile is shown in Fig. 9 and represents the section from z = 1.12 m to z = 1.92 m of a 2.74 m heated length of the driver fuel (CHF is predicted to occur at z  1.75 m during a slow LORA; z is the axial location measured from the start or bottom of the heated length). 2.2. Loop instrumentation Various instruments were installed to measure the following parameters, and the data were recorded using the data acquisition system (DAS).

The experimental approach with descriptions of test trajectories and test procedure has been introduced in Section 1. The planned minimum test section inlet temperature is 18 °C. The nominal pressure at the test section outlet is 158 kPa, while the nominal flow rates are 14.7 kg/s and 9.8 kg/s (corresponding to the analysis of postulated slow LORA and LOFA scenarios, respectively). The nominal flow rate during the low-flow test was revised to 8.8 kg/s (instead of 9.8 kg/s) to include the measurement uncertainty on the low-flow trip (at 9.8 kg/s) of the NRU reactor. Six test trajectories were planned, each for a heated-length and mass flow rate combination. These trajectories are designed on a heat flux vs. local subcooling basis with a number of heat flux levels. Each point of a trajectory represents the local conditions at the downstream end of the bundle simulator heated length. Fig. 3 shows examples of the test trajectories for 0.8-m heatedlength non-uniform AFD bundle at mass flow rates of 14.7 kg/s and 8.8 kg/s. The upper envelope of axial heat flux profiles over the NRU driver fuel rod for the limiting case of a postulated slow loss-of-regulation accident (LORA) scenario is presented in Fig. 3a, and for a loss-of-flow accident (LOFA) scenario in Fig. 3b. The local subcoolings of interest for the NRU driver fuel rod range from 25 K to 85 K at 14.7 kg/s and 23 K to 80 K at 8.8 kg/s (corresponding to the local subcoolings from the heat flux peak location to the downstream end of the fuel rod). In the designed test trajectories for the 0.8-m heated-length non-uniform AFD bundle, the AFD of the simulator at specific powers and inlet conditions is illustrated. The initial contact point of the AFD and the lower-bound curve of the modified Menegus correlation is located near the downstream end of the heated length,

136

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

Fig. 10. Schematic of electrical connection between power supply and heaters.

Table 1 Experimental uncertainty summary. Parameter

Principle

Uncertainty (2r)

Inlet Temperature Outlet Temperature Mass Flow Rate

RTD RTD Turbine

Current Potential Power to FES Power to Bundle Inlet Pressure (PT-1) Outlet Pressure (PT-2) Pressure Drop (DP-1) Pressure Drop (DP-2) Pressure Drop (DP-3) Inlet Subcooling Element Surface Area Flow Area Heat Flux, Uniform AFD Heat Flux, Non-Uniform AFD Velocity

Shunt Direct Voltage Calculated Calculated Absolute Pressure Transducer Absolute Pressure Transducer Differential Pressure Transducer Differential Pressure Transducer Differential Pressure Transducer Calculated Calculated Calculated Calculated Calculated Calculated

0.13 K 0.13 K 0.26% for 14.7 kg/s 0.42% for 8.8 kg/s 0.26% 1.0% 1.0% 0.3% 0.74 kPa 0.42 kPa 0.19 kPa 0.10 kPa 0.81 kPa 0.17 K 0.7% 1.6% 0.8% 3.3% 1.6%

which was obtained by gradually increasing the inlet temperature at constant power. These test trajectories were designed to confirm the adequacy of (i.e., no CHF occurrence below) lower-bound values of the modified Menegus correlation for the NRU driver fuel. The local subcooling was decreased until the occurrence of either CHF or instability in loop operation only during the maximum power test. During the experiment, the outlet pressure was observed to be higher than the nominal pressure (158 kPa) at high flow (14.7 kg/s) and low subcooling test conditions because of the particular combination of loop layout and flow resistance across the plate heat exchanger. Boiling started in the upper plenum resulting in large pressure rises at the test section outlet when local subcooling at -1

the outlet was less than 15 K for the high flow test. The outlet pressure rose to around 300 kPa (which is close to the pressure of 320 kPa predicted at the location of the minimum CHF ratio in the safety analysis for the bounding slow LORA scenario) with increasing power and decreasing subcooling. The impact of a pressure variation on the predicted CHF using the modified Menegus correlation at these conditions is relatively minor as shown in Fig. 11. 4. CHF experiment Three series of tests on NRU driver fuel simulators with three heated lengths and two axial power profiles were completed. 4.1. 1.0-m heated length tests

Mass Flow Rate = 14.7 kg·s 10000

Critical Heat Flux, kW·m-2

9000 8000 7000 6000 5000 4000

P=158 kPa P=200 kPa P=300 kPa

3000

P=180 kPa P=250 kPa

2000 0

20

40

60

80

100

Local Subcooling, K

Fig. 11. Impact of outlet pressure variation on CHF predictions using the modified Menegus correlation at 14.7 kg/s.

The first test series using NRU driver fuel simulators having an axially uniform-heated length of 1 m was performed at the mass flow rates of 14.7 kg/s and 8.8 kg/s. Repeat tests using this bundle type were performed at 14.7 kg/s to examine the test repeatability. All tests were successfully performed following the design test trajectories at all power or heat-flux levels. They confirmed the absence of CHF occurrences at these heat-flux levels along the lower-bound MMC curve. In both high and low flow tests, no CHF occurrences were observed at the maximum heat flux and the minimum achievable local subcooling. The achieved minimum subcooling at the highest heat flux level is 18 K for the 14.7 kg/s tests. During the high power tests with the subcooling approaching 18 K, flashing was initiated inside or just above the outlet plenum (the outlet temperature was close to

137

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

116 °C) as the maximum cooling capability of the loop was reached. This resulted in a significant increase in outlet pressure, making it very difficult to further decrease the local subcooling despite the increase in inlet temperature, and causing the pump inlet temperature to reach its maximum allowable operating value. The test was terminated to prevent possible damage to the test facility. A post-test examination of the simulator showed no visible change (discoloration or deformation) in the bundle over the heated section, compared to the pre-test bundle conditions. The achieved minimum subcooling at the highest heat flux level is 15 K for the 8.8 kg/s tests, beyond which difficulties were encountered in controlling the outlet pressure. 4.2. 0.6-m heated length tests The second test series using NRU driver fuel simulators having an axially uniform-heated length of 0.6 m was performed at the mass flow rates of 14.7 kg/s and 8.8 kg/s. Repeat tests using this bundle type were performed for both high and low flows at high power levels. Tests were performed following the design test trajectories at various power or heat-flux levels. They confirmed the absence of CHF occurrences at these heat-flux levels along the lower-bound MMC curve. In high flow tests, no CHF occurrences were observed at the maximum heat flux and the minimum achievable local subcooling (about 5.74 MW/m2 and 18 K). A valid CHF point was obtained during the repeat tests for 8.8 kg/s. CHF occurred at a power of 1.39 MW with a corresponding outlet coolant subcooling of 39.2 K. After the completion of the experiment, the resistance history of each element was examined. It was found that, between the powers of 1.39 and 1.45 MW, the electrical resistance of Element 2, (which has the least resistance, i.e., the highest power rated element in the bundle) had increased suddenly, indicating the onset of element failure. The photograph of the heaters taken after the tests is shown in Fig. 2a. 4.3. 0.8-m heated length tests The third test series using NRU driver fuel simulators having an axially non-uniform-heated length of 0.8 m was performed at the mass flow rates of 14.7 kg/s and 8.8 kg/s. Repeat tests using this bundle type were performed for the high flow at high power levels. Tests were performed following the design test trajectories and confirmed the absence of CHF occurrences at the tested heat-flux levels along the lower-bound curve. In high flow tests, no CHF occurrences were observed at the maximum power of 1.73 MW with a coolant subcooling of 22.5 K. Loop instability was encountered at subcoolings lower than 22.5 K and the high-flow test was terminated. A valid CHF point was obtained during the tests for 8.8 kg/s. CHF occurred at a power of 1.73 MW when the coolant subcooling reached 20 K. A photograph of the heaters taken after the heater failure (Fig. 2b) shows that the bundle simulator failure occurred initially at (or very close to) the end of the heated length. A resistance history check for each heater suggested that all heaters failed at the same time.3 5. Discussion of CHF results 5.1. Discussion of the combined CHF results from three series of tests CHF results obtained from three series of tests are summarized in Table 2, where the test-section outlet end local conditions at 3

DAS scan rate: 1 scan per second.

maximum presented.

powers

and

maximum

inlet

temperatures

are

5.1.1. Discussion of CHF test results for 14.7 kg/s Fig. 12a4 shows the trajectories for all three series of tests at 14.7 kg/s. To avoid overcrowding, the repeat test results for each series are not shown. Fig. 13a shows the limiting CHF result corresponding to the minimum outlet subcooling at each power lever obtained from three series of tests. The tests for 14.7 kg/s covered 10 average heat flux levels and local subcooling higher than 18 K. No CHF was observed even after exceeding the heat flux corresponding to the lower bound of the modified Menegus correlation by 113%, and after exceeding the modified Menegus correlation by 32%. The maximum heat flux achieved was 5.78 MW/m2, which is 68% higher than the peak value of the upper-bound heat-flux profile encountered in the analysis of the slow LORA scenario. The minimum and maximum achievable local flow subcoolings at the maximum heat flux were 19 K and 107 K5, respectively, covering the range of interest for the NRU driver fuel rod (25–85 K). 5.1.2. Discussion of CHF test results for 8.8 kg/s Fig. 12b shows the trajectories for all three series of tests at 8.8 kg/s. Results of 0.6-m repeat tests, instead of the original tests, are illustrated in Fig. 12b, since a premature CHF was observed during the original tests resulting from the separation of heater material layers. Fig. 13b shows the limiting CHF result corresponding to the minimum outlet subcooling at each power lever obtained from three series of tests. The tests for 8.8 kg/s covered 10 average heat flux levels and local subcooling higher than 15 K. These tests confirmed no CHF occurrence along the lower-bound curve of the modified Menegus correlation and also confirmed no CHF occurrence below and along the curve of the modified Menegus correlation at test conditions. The heat flux corresponding to the minimum subcooling of 15 K was 3.89 MW/m2, which is 106% higher than the prediction of the lower bound of the modified Menegus correlation and 26% higher than the prediction of the modified Menegus correlation. The maximum heat flux achieved was 5.10 MW/m2, which is 125% higher than the peak value of the upper-bound heat-flux profile encountered in the analysis of the LOFA scenario. The maximum achievable local flow subcooling at the maximum heat flux was 95 K6. The local subcoolings of the data covered the range of interest for the NRU driver fuel rod (23–80 K). Two CHF points were obtained, one from 0.6-m uniform AFD tests and another from 0.8-m non-uniform AFD tests. The measured CHF value from 0.6-m uniform AFD tests (5.10 MW/m2) is 99% higher than the prediction of the lower bound of the modified Menegus correlation and 28% higher than the prediction of the modified Menegus correlation with a local subcooling of 39 K. CHF value obtained from 0.8-m non-uniform AFD tests (4.01 MW/m2) is 107% higher than the prediction of the lower bound of the modified Menegus correlation and 28% higher than the prediction of the modified Menegus correlation with a local subcooling of 21 K. 5.1.3. Normalization of the CHF results for 8.8 kg/s The MMC and Lower Bound MMC lines presented in Fig. 13 were obtained based on the nominal values of mass flow rate and pressure (i.e., 14.7 kg/s and 158 kPa) while the test results 4 The MMC and Lower Bound MMC lines presented in Figs. 12–14 were obtained based on the nominal values of mass flow rate and pressure (i.e., 14.7 kg/s and 158 kPa). 5 107 K is the test section inlet subcooling calculated from the recorded data. 6 95 K is the test section inlet subcooling calculated from the recorded data.

138

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

Table 2 CHF results obtained from three test series.a Test series

1.0-m 1.0-m 0.6-m 0.6-m 0.8-m 0.8-m a

14.7 kg/s

Tests Repeat Tests Tests Repeat Tests Tests Repeat Tests

8.8 kg/s

Subcooling at outlet (K)

Heat flux at test section end (kW/m2)

Ratio q00 exp/LB MMC

Ratio q00 exp/MMC

Subcooling at outlet (K)

Heat flux at test section end (kW/m2)

Ratio q00 exp/LB MMC

Ratio q00 exp/MMC

17.9 18.9 18.6 23.7 40.0 22.5

3860 3870 5744 5776 4025 4055

1.37 1.36 2.13 1.98 1.05 1.36

0.85 0.84 1.32 1.24 0.67 0.84

15.5

3885

2.06

1.26

40.1 39.2

4614 5096

1.75 1.99

1.12 1.28

20.5

4045

2.07

1.28

Values presented in the table corresponds to the local conditions at the test section end at the maximum powers and maximum inlet temperatures.

6000

8000 1.0 m Tests MMC

7000

0.6-m Tests Lower Bound MMC

1.0-m Tests MMC

0.8-m Tests AFD Profile LORA

0.6-m Repeat Tests Lower Bound MMC

0.8-m Tests AFD Profile LOFA

5000

Heat Flux, kW·m-2

Heat Flux, kW·m-2

6000 5000 4000 3000

4000

3000

2000

2000

1000 1000

0

0 120

100

80

60

40

20

120

0

100

80

60

40

20

0

Local Subcooling, K

Local Subcooling, K

(a) at 14.7 kg/s

(b) at 8.8 kg/s

Fig. 12. CHF results obtained from three series of tests.

8000

1.0 m Tests MMC

7000

0.6-m Tests Lower Bound MMC

6000

0.8-m Tests AFD Profile LORA

1.0-m Tests CHF Points AFD Profile LOFA

0.6-m Repeat Tests MMC

0.8-m Tests Lower Bound MMC

5000

Heat Flux, kW·m-2

Heat Flux, kW·m-2

6000 5000 4000 3000

4000

3000

2000

2000 1000 1000 0

0 40

35

30

25

20

15

10

Local Subcooling, K

60

50

40

30

20

10

Local Subcooling, K

(b) at 8.8 kg/s

(a) at 14.7 kg/s Fig. 13. Limiting CHF test results obtained from three series of tests.

correspond to the measured mass flow rates and pressures. To make a more accurate comparison between the measured and predicted CHF values, the measured CHF values were normalized with respect to deviations from the nominal mass flow rate and nominal pressure.7 The normalized CHF value at the nominal mass flow rate _ nor Þ and nominal pressure ðPnor Þ, CHF nor Pnor ;m_ nor , was calculated as ðm

CHF nor

_ nor P nor ;m

¼ CHF exp þ

_ exp Pexp ;m

þ

@CHF ðPnor  Pexp Þ @P

@CHF _ nor  m _ exp Þ ðm _ @m

ð1Þ

7 CHF increases with increasing mass flow rate and increasing pressure at low subcoolings.

where CHF exp Pexp ;m_ exp is the experimental CHF value at the measured _ exp Þ and measured pressure ðP exp Þ. The derivatives, mass flow rate ðm @CHF @CHF and , can be estimated based on the modified Menegus cor_ @P @m relation by applying a correction factor of 1.28 (since the measured CHF is 28% higher than the prediction of the modified Menegus correlation). The calculated normalized CHF values obtained from 0.6-m and 0.8-m tests are presented in Fig. 14. For comparison, the normalized limiting data point corresponding to the minimum outlet subcooling at maximum power level obtained from 1.0-m tests (26% above MMC) is also presented in Fig. 14. All these data points follow a consistent trend with local subcooling when compared to the modified Menegus correlation.

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140 6000

CHF Point, 0.6-m Limiting Point, 1.0-m Lower Bound MMC

Heat Flux, kW·m-2

5000

CHF Point, 0.8-m MMC AFD Profile LOFA

4000

3000

2000

1000

0 50

45

40

35

30

25

20

15

10

Local Subcooling, K

Fig. 14. Normalized CHF test results obtained from three series of tests, mass flow rate at 8.8 kg/s.

5.2. Effect of wall thickness tolerance on CHF As described in Section 2, all bundle simulators simulate the radial power profile corresponding to the fresh fuel. The ratio of element power to bundle-average element power is 1.031 for the outer ring elements and 0.906 for the inner ring elements. Ideally, the electrical resistances of the heaters and the hence the powers to the heaters in the same ring should be the same resulting in a simultaneous CHF occurrence in outer-ring heaters.8 However, due to the machining tolerance, the wall thickness and hence the resistance of each heater in the same ring would not be the same. CHF would initially occur on the outer-ring heater with the least resistance (i.e., the highest rated heater) if the difference in resistance were large. Therefore, the measured CHF value would be lower than that corresponding to the bundle consisting of heaters with identical resistance in each ring. In other words, the measured CHF is conservative. In the 0.8-m non-uniform AFD bundle tests, CHF occurred on each heater at almost the same time (in one second) since the difference in heater resistance of the outer ring is relatively small (the maximum heater-to-ring-average resistance ratio is 1.016). The local heat flux of the highest rated heater at CHF location is 112% higher than the prediction of the lower bound of the modified Menegus correlation and 29% higher than the prediction of the modified Menegus correlation. For the 0.6-m uniform AFD bundle, however, the resistance difference among the outer-ring heaters is larger (the maximum heater-to-ring-average resistance ratio is 1.031), hence CHF initially occurred on the heater with the least resistance (i.e., the highest rated heater). The local heat flux of the highest rated heater at CHF conditions is 106% higher than the prediction of the lower bound of the modified Menegus correlation and 30% higher than the prediction of the modified Menegus correlation.

5.3. Effect of axial power profile on CHF Bundle simulators having the 1.0-m and 0.6-m heated lengths were designed with uniform axial power profile to expedite the experiment (manufacturing of FESs with a non-uniform axial power profile is more complex and time-consuming). A separate simulator having the 0.8-m heated length with a non-uniform axial power profile was constructed to provide confirmatory data for the CHF prediction. 8 CHF would initially occur in the outer ring heaters since the local heat fluxes of the outer-ring heaters are much higher than those of the inner-ring heaters (14%).

139

Previously studies on the effect of axial power profile on CHF (Yang et al., 2006) suggested that the axial power profile has a strong effect on CHF at high qualities where the CHF mechanism is liquid film dryout in an annular flow regime. The axial power profile effect on CHF diminishes with decreasing dryout quality. At low dryout qualities (less than 0.1), the effect of axial power profile on CHF is small. For the departure from nucleate boiling (DNB) type of CHF, especially for that corresponding to the mechanism of micro-layer evaporation from a nucleation site at high subcoolings, the critical condition is basically induced by local overheating. Thus, the local heat flux controls the subcooled CHF and the axial power profile effect becomes negligible. The CHF mechanism corresponds to departure from nucleate boiling for the NRU driver fuel since CHF occurred at subcooled conditions (see Figs. 12 and 13). It is independent from the axial power profile (i.e., local heat flux is the dominant factor). The CHF experimental results support this statement. The measured CHF value from the 0.6-m uniform power profile tests is 99% higher than the prediction of the lower bound of the modified Menegus correlation and 28% higher than the prediction of the modified Menegus correlation. CHF value obtained from non-uniform power profile tests is 107% higher than the prediction of the lower bound of the modified Menegus correlation and 28% higher than the prediction of the modified Menegus correlation. Considering also the wall thickness tolerance and the uncertainty of the local heat flux (0.8% for uniform power profile and 3.3% for the non-uniform power profile), the CHF values for the uniform and non-uniform power profiles follow a consistent trend with local subcooling when compared to the modified Menegus correlation. 5.4. Effect of low-impact spacer on CHF To maintain the correct element spacing, actual NRU grid spacers were used (each spacer occupies 29.3% of the undisturbed cross-sectional flow area) and placed at regular intervals (476.3 mm). In addition, low-impact spacers were introduced midway between adjacent NRU spacer locations to prevent the bundle simulator from potential deforming due to large magnetic forces. Each of these low-impact spacers occupies 11.5% of the undisturbed cross-sectional flow area (as compared to 29.3% for the NRU spacer) and has rounded leading and trailing edges. The distance between the most downstream low-impact spacer and the downstream end of the bundle heated length (CHF location) is 188 mm. A previous experimental study (Yang et al., 2005) showed that the effect of flow obstruction on CHF strongly depends on the flow obstruction configuration, i.e., the flow area blockage and the distance between two adjacent obstacles (obstacle pitch). CHF occurred at the location just before the obstacle near the downstream end of the heated length. Large CHF enhancement was observed for large flow blockages with short obstacle pitch. The CHF enhancement decreases with a decrease in dryout quality, a decrease in flow blockage, an increase in mass flux and an increase in flow obstruction pitch. Fig. 15a and b show comparisons of CHF and critical power enhancements between obstacles with large and small flow blockages at the same obstacle pitch of 160 mm. It suggests that the effect of flow obstruction on CHF and critical power for obstacles with 10% flow blockage at an obstacle pitch of 160 mm are negligible at low qualities (dryout quality less than 0.1). Since the current tests were performed at subcooled conditions (i.e., negative qualities at the downstream end of heated length), and the flow blockage of the low-impact spacer (11.5%) is close to 10%, and the distance between the most downstream low-impact spacer and the CHF location is larger than 160 mm, the effect of the low-impact spacer on CHF and critical power is expected to be negligible.

140

J. Yang et al. / Nuclear Engineering and Design 313 (2017) 129–140

(a) CHF vs. Dryout Quality

(b) CHF Power vs. Inlet Quality

Fig. 15. Comparisons of CHF vs. dryout quality and CHF power vs. inlet quality for uniform AFD bare tubes with and without flow obstruction (Yang et al.., 2005). Uniform bare: bare tube with uniform AFD. Uniform OB, 20%  160 mm: uniform AFD tube with cylindrical obstacles of 20% flow area blockage and 160 mm pitch. Uniform OB, 10%  160 mm: uniform AFD tube with cylindrical obstacles of 10% flow area blockage and 160 mm pitch.

6. Conclusions and final remarks The NRU driver fuel assembly consists of twelve finned fuel elements and is operated at highly subcooled conditions. CHF experiment for a finned element bundle at such conditions is very difficult, costly and time consuming since the CHF mechanism is the departure from nucleate boiling, for which physical burnout is expected when CHF is reached. Therefore, the relevant experimental information is very limited in the open literature. A novel experimental approach was developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the occurrence of this type of CHF. Consequently the experiment was successfully conducted based on the developed approach with the experimental data (Table 2) and analysis results presented in present paper. CHF tests have been performed at two mass flows (14.7 kg/s and 8.8 kg/s) using NRU driver fuel simulators having three different heated lengths: 1.0 m and 0.6 m with a uniform axial power profile and 0.8 m with a non-uniform axial power profile. The experimental results confirmed that the current CHF prediction method predicts lower CHF values than experimentally measured using the NRU driver fuel rod simulators. 1. The tests confirmed no CHF occurrence along the lower-bound curve of the modified Menegus correlation and also confirmed no CHF occurrence along the curve of the modified Menegus correlation at test conditions. 2. The tests for 14.7 kg/s covered 10 average heat flux levels and local subcoolings higher than 18 K (which bounds the range of interest for the analysis of the postulated slow LORA scenario). The tests for 8.8 kg/s covered 10 average heat flux levels and local subcoolings higher than 12 K (which bounds the range of interest for the analysis of the postulated LOFA scenario). 3. At a flow of 14.7 kg/s, the maximum heat flux achieved was 5.78 MW/m2, which is 68% higher than the peak value of the upper-bound heat-flux profile encountered in the analysis of the slow LORA scenario. At a flow of 8.8 kg/s, the maximum heat flux achieved was 5.10 MW/m2, which is 125% higher than the peak value of the upper-bound heat-flux profile encountered in the analysis of the LOFA scenario. 4. At a flow of 14.7 kg/s, no CHF was observed even after exceeding the heat flux corresponding to the lower bound of the modified Menegus correlation by 113%, and after exceeding the

modified Menegus correlation by 32%. At a flow of 8.8 kg/s, two CHF points were obtained, one from 0.6-m uniform AFD tests and another from 0.8-m non-uniform AFD tests. The experimental CHF value from 0.6-m uniform AFD tests (5.10 MW/m2) is 99% higher than the prediction of the lower bound of the modified Menegus correlation and 28% higher than the prediction of the modified Menegus correlation, at local subcooling of 39 K. The CHF value obtained from 0.8-m non-uniform AFD tests (4.01 MW/m2) is 107% higher than the prediction of the lower bound of the modified Menegus correlation and 28% higher than the prediction of the modified Menegus correlation, at local subcooling of 21 K. These two CHF points follow a consistent trend with local subcooling. 5. The use of low-impact spacers in the NRU driver fuel simulators has no effect on CHF and has small effect on the bundle hydraulic characteristics. Acknowledgments The successful completion of this project would not have been possible without the significant effort of many people. In particular the lab supervisor Y. Lachance and the lead technologist B. Addicott played a key role in the timely completion of this project. The valuable contributions to the experiment from J. Romain, M. Dickerson, V. Gauthier, E. Lux, W. McGillis, C. Paulusse and J. Sheedy are recognized. The technical inputs from Y. Guo, and the review comments from L.K.H. Leung and A. Vasic are hereby acknowledged. References Menegus, R.L., 1959. Burnout of Heating Surfaces in Water, duPont de Nemours Report DP-363. Yang, J., Groeneveld, D.C., Leung, L.K.H., Cheng, S.C., El Nakla, M.A., 2006. An experimental and analytical study of the effect of axial power profile on CHF. Nucl. Eng. Des. 236 (13), 1384–1395. Yang, J., Yuan, L.Q., Shan, J.Q., Cheng, S.C. and Groeneveld, D.C., 2005. An Investigation of the Impact of Flow Obstruction on the Effect of Axial HeatFlux Distribution on CHF in Tubes, University of Ottawa Report, UO-MCG-TH2005-001. Jun Yang has been working at the Canadian Nuclear Laboratories since 2005 as a research scientist in thermalhydraulics Branch. He is a section head leading various thermalhydraulics experimental and analytical activities.