*2.1.2 The validation procedures*

The Validation Process of a thermal–hydraulic system code calculation has the goal to demonstrate that the code results (obtained by the application of the code with the developed nodalization) constitute a realistic approximation of the reference plant behavior (a full-size Nuclear Power Plant or a facility). The flow chart of the adopted Validation Process is given in **Figure 3**.

A SM representing an actual system (ITF or NPP) is qualified when:


Based on this, three main phases of the Validation Process can be distinguished:


In relation to the first step of the methodology it is worth demonstrating that the discrepancies between relevant geometrical parameters of the plant and the data implemented into the nodalization are within acceptable values.

The second step of the Validation Process deals with the achievement of the steady state. A set of significant parameters is identified to demonstrate that the discrepancies between calculated and measured data available from nominal stationary conditions are within acceptability thresholds.

**247**

**Figure 4.**

*International Benchmark Activity in the Field of Sodium Fast Reactors*

The third step of the Validation Process is the "On-transient" validation, a very complex step requiring several different sub-steps which include qualitative and quantitative accuracy evaluations performed to evaluate the acceptability of the calculation on "transient level". If the qualitative accuracy evaluation is acceptable, the accuracy of the code calculation can be quantified utilizing the Fast Fourier

The EBR-II plant, located in Idaho, was operated by ANL for the U.S.

Department of Energy from the beginning of 1964 until 1994. EBR-II rated thermal power was 62.5 MW, with electric output of 20MW. EBR-II was a sodium-cooled reactor fueled with uranium metal alloy fuel, with a pool type primary system. **Figure 4** shows the configuration of the main components in the EBR-II primary

All major primary system components were submerged in the primary tank. Two primary pumps drew sodium from the pool and provided sodium to the two inlet plena for the core, through high pressure and low-pressure pipes. The reactor

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

Transform Based Method (FFTBM) [11].

*Flow Chart of the Validation Procedures for the SM.*

**Figure 3.**

**2.2 EBR-II plant and the developed SM**

system [12] together with the developed RELAP5 SM.

*(a) EBR-II Primary System and (b) The adopted Nodalization.*

*International Benchmark Activity in the Field of Sodium Fast Reactors DOI: http://dx.doi.org/10.5772/intechopen.97812*

#### **Figure 3.**

*Recent Advances in Numerical Simulations*

The flow-chart linking the RDS, the Input deck, the VR and the EH is highlighted in **Figure 2**. The solid lines show the time sequence of the activities, the dotted lines indicate the feedback for the review and the dashed lines are the

The Validation Process of a thermal–hydraulic system code calculation has the goal to demonstrate that the code results (obtained by the application of the code with the developed nodalization) constitute a realistic approximation of the reference plant behavior (a full-size Nuclear Power Plant or a facility). The flow chart of

A SM representing an actual system (ITF or NPP) is qualified when:

• It reproduces the measured nominal steady state condition of the system;

Based on this, three main phases of the Validation Process can be distinguished:

1.The demonstration of the geometrical fidelity of the developed nodalization;

In relation to the first step of the methodology it is worth demonstrating that the discrepancies between relevant geometrical parameters of the plant and the data

The second step of the Validation Process deals with the achievement of the steady state. A set of significant parameters is identified to demonstrate that the discrepancies between calculated and measured data available from nominal

• It enjoys a large geometrical fidelity with the involved system;

• It shows a satisfactory behavior in time dependent conditions.

2.The demonstration of the steady state achievement;

implemented into the nodalization are within acceptable values.

stationary conditions are within acceptability thresholds.

3.The "on-transient" Validation.

necessary input to develop the input deck and the EH.

the adopted Validation Process is given in **Figure 3**.

*2.1.2 The validation procedures*

**Figure 2.**

*SCCRED Flow Chart.*

**246**

*Flow Chart of the Validation Procedures for the SM.*

The third step of the Validation Process is the "On-transient" validation, a very complex step requiring several different sub-steps which include qualitative and quantitative accuracy evaluations performed to evaluate the acceptability of the calculation on "transient level". If the qualitative accuracy evaluation is acceptable, the accuracy of the code calculation can be quantified utilizing the Fast Fourier Transform Based Method (FFTBM) [11].

### **2.2 EBR-II plant and the developed SM**

The EBR-II plant, located in Idaho, was operated by ANL for the U.S. Department of Energy from the beginning of 1964 until 1994. EBR-II rated thermal power was 62.5 MW, with electric output of 20MW. EBR-II was a sodium-cooled reactor fueled with uranium metal alloy fuel, with a pool type primary system. **Figure 4** shows the configuration of the main components in the EBR-II primary system [12] together with the developed RELAP5 SM.

All major primary system components were submerged in the primary tank. Two primary pumps drew sodium from the pool and provided sodium to the two inlet plena for the core, through high pressure and low-pressure pipes. The reactor

**Figure 4.** *(a) EBR-II Primary System and (b) The adopted Nodalization.*

vessel accommodated 637 hexagonal subassemblies divided in three regions: central core (up to row 5), inner blanket (rows 6 and 7) and outer blanket (up to row 16). Hot sodium exited the subassemblies into a common upper plenum where it mixed before passing through the reactor outlet pipe ("Z-pipe") into the Intermediate Heat Exchanger (IHX). Sodium then exited the IHX into the primary sodium tank before entering sodium primary pumps.

The EBR-II benchmark specifications [12] were used to develop a detailed thermal hydraulic model (see **Figure 4b**) of the reactor. The RELAP5 system thermal hydraulic code was used for both performing the nodalization and running the calculations.

The whole reactor core consisted of 96 channels representing all 10 types of subassemblies used in the reactor, and two bypass flow paths. The reactor vessel was first subdivided into 16 rows. The subassemblies in the first 6 rows have been modeled separately (1 by 1) with 81 channels, except for safety/control rods that have been merged into one channel. Rows 7 to 16 made of reflector and blanket subassemblies have been modeled with one channel per type of subassembly in each row. One heat structure component has been used to simulate the active part of the fuel pins for each subassembly in the central core region, assuming a flat and constant power profile along all the active length. The pool was modeled with a cylindrical multi-dimensional component having 3 azimuthal meshes, two of which were thermally linked to the pumps and the third one to the IHX. The heat exchanger was of counter-current flow type. The primary side of IHX has been modeled with a pipe which takes hot sodium flowing out from the "Z-pipe" and discharges the cold sodium directly into the pool. The intermediate side of IHX has also been modeled with a pipe equivalent to 3026 secondary tubes through which the intermediate sodium flows. The boundary conditions for the intermediate side were imposed by the time-dependent volume and time-dependent junction components.

#### **2.3 Transient results and sensitivity analysis of EBR-II SM**

#### *2.3.1 Reference results*

The transient was initiated by a trip of primary and intermediate pumps, which instantaneously scrammed the reactor. While the coast-down shapes for SHRT-17 were designed to be identical for the two primary pumps, intrinsic differences between the two pump drive units caused a difference in the stop times.

The transient calculation was performed after the achievement of acceptable steady-state conditions. Starting from full power and flow, both the primary loop and the intermediate loop coolant pumps were simultaneously tripped, and the reactor was scrammed to simulate a protected loss-of-flow accident. Therefore, in the early stage of the pump coast-down (up to about 10 s) the cladding and the outlet coolant temperature decreased. During the transition from the forced to natural circulation (between 10 and 100 s) the mass flow rates decreased rapidly and the unbalance between the total core power and the energy removed from the primary coolant caused a rapid increase of the cladding temperatures and a slower increase of the coolant temperatures. When the natural circulation is fully established (after about 100 s) the total core power is efficiently removed in all subassemblies and the coolant and cladding temperatures decrease.

During the pump coast-down the mass flow rate in the instrumented subassembly XX09 remained a little bit higher than the experimental data (see **Figure 5**), which affected the coolant and cladding temperatures in the whole subassembly. Indeed, both the coolant temperatures below (**Figure 6**) and above (**Figure 7**) the active core region and the cladding temperatures at the middle and at the top of

**249**

simulation.

**Figure 7.**

**Figure 5.**

**Figure 6.**

*XX09 Lower and Upper Flowmeter Coolant Temperatures.*

*XX09 Outlet and Thimble Annulus Coolant Temperatures.*

*XX09 Mass Flow Rate.*

*International Benchmark Activity in the Field of Sodium Fast Reactors*

the core (**Figure 8**) were slightly lower than the experimental data. These small differences became negligible during the long-term cooling because the mass flow rate reached the correct value. It should be noted that the flowmeter temperatures, where the gamma heating occurs, were qualitatively correctly predicted by the

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

*International Benchmark Activity in the Field of Sodium Fast Reactors DOI: http://dx.doi.org/10.5772/intechopen.97812*

**Figure 5.** *XX09 Mass Flow Rate.*

*Recent Advances in Numerical Simulations*

before entering sodium primary pumps.

calculations.

*2.3.1 Reference results*

vessel accommodated 637 hexagonal subassemblies divided in three regions: central core (up to row 5), inner blanket (rows 6 and 7) and outer blanket (up to row 16). Hot sodium exited the subassemblies into a common upper plenum where it mixed before passing through the reactor outlet pipe ("Z-pipe") into the Intermediate Heat Exchanger (IHX). Sodium then exited the IHX into the primary sodium tank

The EBR-II benchmark specifications [12] were used to develop a detailed thermal hydraulic model (see **Figure 4b**) of the reactor. The RELAP5 system thermal hydraulic code was used for both performing the nodalization and running the

The whole reactor core consisted of 96 channels representing all 10 types of subassemblies used in the reactor, and two bypass flow paths. The reactor vessel was first subdivided into 16 rows. The subassemblies in the first 6 rows have been modeled separately (1 by 1) with 81 channels, except for safety/control rods that have been merged into one channel. Rows 7 to 16 made of reflector and blanket subassemblies have been modeled with one channel per type of subassembly in each row. One heat structure component has been used to simulate the active part of the fuel pins for each subassembly in the central core region, assuming a flat and constant power profile along all the active length. The pool was modeled with a cylindrical multi-dimensional component having 3 azimuthal meshes, two of which were thermally linked to the pumps and the third one to the IHX. The heat exchanger was of counter-current flow type. The primary side of IHX has been modeled with a pipe which takes hot sodium flowing out from the "Z-pipe" and discharges the cold sodium directly into the pool. The intermediate side of IHX has also been modeled with a pipe equivalent to 3026 secondary tubes through which the intermediate sodium flows. The boundary conditions for the intermediate side were imposed by

the time-dependent volume and time-dependent junction components.

The transient was initiated by a trip of primary and intermediate pumps, which instantaneously scrammed the reactor. While the coast-down shapes for SHRT-17 were designed to be identical for the two primary pumps, intrinsic differences between the two pump drive units caused a difference in the stop times.

The transient calculation was performed after the achievement of acceptable steady-state conditions. Starting from full power and flow, both the primary loop and the intermediate loop coolant pumps were simultaneously tripped, and the reactor was scrammed to simulate a protected loss-of-flow accident. Therefore, in the early stage of the pump coast-down (up to about 10 s) the cladding and the outlet coolant temperature decreased. During the transition from the forced to natural circulation (between 10 and 100 s) the mass flow rates decreased rapidly and the unbalance between the total core power and the energy removed from the primary coolant caused a rapid increase of the cladding temperatures and a slower increase of the coolant temperatures. When the natural circulation is fully established (after about 100 s) the total core power is efficiently removed in all subassemblies and the

During the pump coast-down the mass flow rate in the instrumented subassembly XX09 remained a little bit higher than the experimental data (see **Figure 5**), which affected the coolant and cladding temperatures in the whole subassembly. Indeed, both the coolant temperatures below (**Figure 6**) and above (**Figure 7**) the active core region and the cladding temperatures at the middle and at the top of

**2.3 Transient results and sensitivity analysis of EBR-II SM**

coolant and cladding temperatures decrease.

**248**

**Figure 6.** *XX09 Lower and Upper Flowmeter Coolant Temperatures.*

#### **Figure 7.**

*XX09 Outlet and Thimble Annulus Coolant Temperatures.*

the core (**Figure 8**) were slightly lower than the experimental data. These small differences became negligible during the long-term cooling because the mass flow rate reached the correct value. It should be noted that the flowmeter temperatures, where the gamma heating occurs, were qualitatively correctly predicted by the simulation.

**Figure 8.** *XX09 Clad Temperatures.*

#### *2.3.2 Sensitivity analysis*

During the phase 2 of the benchmark, a sensitivity analysis on the gamma heating was performed aimed at understanding the experimental behavior of the coolant temperature at the inlet and outlet of instrumented subassembly (in particular, the instrumented subassembly XX09).

To perform the sensitivity analysis a simple model (see **Figure 9**) of the instrumented subassembly XX09 was developed considering only the subassembly channel and the guide thimble annulus channel, thermally connected with a passive heat structure simulating the subassembly walls. The heat structure simulating the guide thimble wall has been isolated. Regarding the active heat structure, in addition to the flat power profile adopted in the early stage of the phase 2 of benchmark (Phase-2A), four different axial power distribution (see **Figure 9**) have been implemented:


**251**

**Figure 10.**

*Upper Flowmeter Temperature.*

*International Benchmark Activity in the Field of Sodium Fast Reactors*

the upper flowmeter thermocouples (see **Figure 10**).

application of the Validation Process are available in [7].

experimental data were within acceptability thresholds.

fidelity between the EBR-II hardware and the developed nodalization.

The axial power distribution below the BAF has been calculated to match the experimental steady state values of the coolant temperature at the lower and upper flowmeter thermocouples. It can be noted that the power supplied below the active part of the core (sensitivity #1, 3 and 4) positively affects the temperature trends at

On the contrary, the power supplied above the active part of the core (sensitivity

#2, 3 and 4) results in minor effect on the temperature trends. In particular, the coolant outlet temperature (see **Figure 11**) shows a light delay in the temperature increase after the pump coastdown compared to the experimental data and to the

In the framework of the benchmark, a simplified version of the Validation Process was adopted selecting a smaller set of parameters to carry-out the demonstration of the geometrical fidelity and the demonstration of the steady state achievement (see §2.1.2). In addition, only a quantitative analysis by the FFTBM was carried-out, without performing the qualitative analysis (which is instead a mandatory step for a full application of the Validation Process) due to limited project's recourses. The main goal of the quantitative evaluation, as well as the analysis carried out, was to support the interpretation of the results calculated by the CRP participants, i.e., to provide quantitative measures of the discrepancies between the assumptions made by the participants and the reference specification data. These discrepancies can provide a support to understand the reasons for the differences between the participants' results and the experimental data. Results from the

First, a list of more than 50 parameters was selected to perform the geometrical

For the achievement of the steady state, a set of significant parameters was identified to demonstrate that the discrepancies between calculated and measured

Regarding the third step of the Validation Process, the "On-transient" Validation, the focus was only on the so called "Quantitative Accuracy Evaluation" that is performed by the FFTBM. A list of about 50 parameters was selected,

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

**2.4 Validation process of the EBR-II SM**

other sensitivity cases.

**Figure 9.** *XX09 Model and Axial LHR used in the Sensitivity Analysis.*

*International Benchmark Activity in the Field of Sodium Fast Reactors DOI: http://dx.doi.org/10.5772/intechopen.97812*

The axial power distribution below the BAF has been calculated to match the experimental steady state values of the coolant temperature at the lower and upper flowmeter thermocouples. It can be noted that the power supplied below the active part of the core (sensitivity #1, 3 and 4) positively affects the temperature trends at the upper flowmeter thermocouples (see **Figure 10**).

On the contrary, the power supplied above the active part of the core (sensitivity #2, 3 and 4) results in minor effect on the temperature trends. In particular, the coolant outlet temperature (see **Figure 11**) shows a light delay in the temperature increase after the pump coastdown compared to the experimental data and to the other sensitivity cases.

### **2.4 Validation process of the EBR-II SM**

*Recent Advances in Numerical Simulations*

*2.3.2 Sensitivity analysis*

*XX09 Clad Temperatures.*

**Figure 8.**

implemented:

the instrumented subassembly XX09).

During the phase 2 of the benchmark, a sensitivity analysis on the gamma heating was performed aimed at understanding the experimental behavior of the coolant temperature at the inlet and outlet of instrumented subassembly (in particular,

To perform the sensitivity analysis a simple model (see **Figure 9**) of the instru-

mented subassembly XX09 was developed considering only the subassembly channel and the guide thimble annulus channel, thermally connected with a passive heat structure simulating the subassembly walls. The heat structure simulating the guide thimble wall has been isolated. Regarding the active heat structure, in addition to the flat power profile adopted in the early stage of the phase 2 of benchmark (Phase-2A), four different axial power distribution (see **Figure 9**) have been

1.Power supplied also below the active part of the core;

2.Power supplied also above the active part of the core;

4.Axial power distribution as in SHRT-45.

*XX09 Model and Axial LHR used in the Sensitivity Analysis.*

3.Power supplied also above and below the active part of the core;

**250**

**Figure 9.**

In the framework of the benchmark, a simplified version of the Validation Process was adopted selecting a smaller set of parameters to carry-out the demonstration of the geometrical fidelity and the demonstration of the steady state achievement (see §2.1.2). In addition, only a quantitative analysis by the FFTBM was carried-out, without performing the qualitative analysis (which is instead a mandatory step for a full application of the Validation Process) due to limited project's recourses. The main goal of the quantitative evaluation, as well as the analysis carried out, was to support the interpretation of the results calculated by the CRP participants, i.e., to provide quantitative measures of the discrepancies between the assumptions made by the participants and the reference specification data. These discrepancies can provide a support to understand the reasons for the differences between the participants' results and the experimental data. Results from the application of the Validation Process are available in [7].

First, a list of more than 50 parameters was selected to perform the geometrical fidelity between the EBR-II hardware and the developed nodalization.

For the achievement of the steady state, a set of significant parameters was identified to demonstrate that the discrepancies between calculated and measured experimental data were within acceptability thresholds.

Regarding the third step of the Validation Process, the "On-transient" Validation, the focus was only on the so called "Quantitative Accuracy Evaluation" that is performed by the FFTBM. A list of about 50 parameters was selected,

**Figure 10.** *Upper Flowmeter Temperature.*

**Figure 11.** *Coolant Outlet Temperature.*

varying from power, absolute pressures, velocity and mass flow rates, fluid temperatures, rod surface temperatures, pressure drops and mass inventory. In the case of parameters for which no reference or measured value was available a code-to-code comparison was performed.
