**3.1 Overview of FFTF and of the developed SM**

*Recent Advances in Numerical Simulations*

comparison was performed.

*Coolant Outlet Temperature.*

passive safety.

**Figure 11.**

operating until 1992 [13].

**3. Analysis of FFTF LOFWOS Test #13**

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

The IAEA CRP focused on benchmark analysis of one of the unprotected passive

safety demonstration tests performed at the Fast Flux Test Facility (FFTF) was launched in 2018 to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of sodium fast reactor

The Fast Flux Test Facility was a 400 MW-thermal loop type SFR prototype with mixed oxide fuel, built to assist development and testing of advanced fuels and materials for fast breeder reactors. It was located at the Hanford site in Washington and designed by the Westinghouse Electric Corporation for the U.S. Department of Energy (DOE). FFTF reached criticality in 1980 and has been

The loss of flow without scram (LOFWOS) Test #13 was performed on July 18, 1986 as part of the Passive Safety Testing (PST) program with the aim of confirming the safety margins of FFTF as a liquid metal reactor, providing data for computer code validation, and demonstrating the inherent and passive safety benefits of its specific design features. One of the passive reactivity control devices are the Gas Expansion Modules (GEMs) located at the periphery of the FFTF core. GEMs are hollow tubes sealed at the top and open on the bottom with Argon cover gas trapped inside. During normal operation, the pressure head of the primary pumps compresses the gas to a level above the active part of the core, filling the GEMs with sodium. Following a pump trip and a corresponding decrease in the sodium pressure, the trapped gas would expand and displace sodium, increasing the neutron leakage from the core and decreasing the core

Starting from 50% power and 100% flow, the Test #13 was initiated when the three primary sodium pumps were simultaneously tripped. The secondary loop

sodium pumps remained operational throughout the whole test.

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reactivity.

An overview of the FFTF coolant system is shown in **Figure 12**, where three main parts can be distinguished: the reactor vessel, the primary loop, and the secondary loop.

Regarding the reactor vessel, cold sodium was discharged from the three primary loop inlet pipes into an inlet plenum at the bottom of the reactor vessel. Sodium was then drawn up into the core support structure and distributed to the core assemblies and radial shields, as well as leakage and bypass flow paths. Sodium discharged from these flow paths was mixed above a horizontal baffle plate in a common outlet plenum before exiting the reactor vessel through one of three primary loop outlet pipes. The outlet plenum was bounded at the top by a region of Argon cover gas.

The IHX was vertically mounted counterflow shell and tube designs and separated activated sodium coolant in the primary loops from nonradioactive sodium in the secondary loops. Within each secondary loop, hot leg piping ran from the IHX outlet to a Dump Heat Exchanger (DHX) unit, which discharged heat to the environment. Each DHX unit contained four individual sodium-to-air dump heat exchanger modules. The cold leg sodium ran from the DHX unit to a sodium pump, and back to the IHX.

The FFTF core was loaded with199 hexagonal assemblies, that could be grouped in 91 core locations, from row 1 to row 6, including 7 different types of driver fuel assemblies, control and safety rods and test locations, 60 internal reflector assemblies, in rows 7 and 8A, and 48 external reflector assemblies, in rows 8B and 9.

Starting from the benchmark specifications [13], a detailed SM reproducing each component depicted in **Figure 12** was developed following the NINE nodalization techniques, except for the DHXs, that were replaced by boundary conditions.

The reactor vessel has been modeled with a cylindrical multi-dimensional component having three radial meshes: the innermost region represents the area occupied by the core basket and the leakage flow that passes around the fuel assemblies and reflector assemblies (up to row 8A), the intermediate zone models the annular plenum and the flow around reflector assemblies (rows 8B and 9) and through radial shields, and the outermost region simulates the peripheral plenum and the in-vessel storage region. The flow through the assemblies in the reactor core was modeled with 18 channels, 16 pipe components simulating the sixteen assembly flow zones representing the different types of assemblies and 2 pipe components to simulate separately the instrumented assemblies, the Row 2 fast response Proximity

**Figure 12.** *FFTF Coolant System Overview.*

Instrumented Open Test Assemblies (PIOTA) and the Row 6 fast response PIOTA. As for the hydraulic part, one heat structure was inserted in each channel to simulate the active part of the assemblies. A flat axial power profile was imposed along the active length of all the assemblies.

The three primary loops and secondary loops were modeled separately, each one having the same number of hydraulic components. Regarding the secondary loops, two time-dependent components were inserted to provide the proper boundary conditions, one component at the exit of the hot leg piping to set the secondary side pressure, and one component at the beginning of the cold leg piping, downstream the DHXs, to set the appropriate sodium temperature.

### **3.2 Reference results and sensitivity analysis of the FFTF SM**

After imposing the boundary conditions (i.e., pumps speed, core power and secondary loops flow conditions) provided by the benchmark team, acceptable steady-state conditions were achieved before performing the transient simulation. The RELAP5 system thermal hydraulic code was used to make the analysis.

The FFTF LOFWOS Test #13 was initiated when the primary pumps tripped simultaneously. The row 2 PIOTA outlet temperature shown in **Figure 13** can be observed to describe the behavior of the FFTF core during the transient.

The initial rapid rise of the PIOTA outlet temperature was caused by the increasing core power-to-flow ratio following the pump trips. The outlet temperature peaked at around 10 seconds when the power-to-flow ratio reached its maximum value. Then, the increase in core temperatures together with the drop of the GEM sodium level introduced a large negative reactivity feedback, so power decreased faster than the primary flow rate. The drop in reactor power was quick enough to compensate for the reduced flow rate in the primary loop and the sodium temperature started to decrease. As the GEM sodium level approached the bottom of the core, the negative reactivity insertion slowed down. The core outlet temperature began to rise again, and a second peak occurred when the natural circulation was established. Natural circulation was maintained while power continued to decrease resulting in a decrease of the core outlet temperature until the end of the test.

**Figure 13** shows the comparison of the PIOTA outlet temperature predicted by the SM against the measurement (in this and subsequent figures the solid line shows the experimental data while the dashed line displays the SM results). The SM results are in good agreement with the experimental data for the entire duration of

**255**

**Figure 14.**

*Primary Loop Hot and Cold Leg Temperatures.*

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

the transient. In particular, the time of occurrence of the two peaks is captured very

The cold leg and hot leg temperatures in one of the primary loops are shown in **Figure 14**. The hot leg fluid temperature has been quite well predicted by the SM, showing a trend slightly oscillating around the experimental value. That may be

different prediction of the sodium mixing and thermal stratification phenomena in the outlet plenum of the reactor vessel during the natural circulation phase that are difficult to simulate by a system thermal–hydraulic code. The calculated cold leg fluid temperature showed a faster rise at the beginning of the transient following the increase in the DHX sodium outlet temperatures, it reached a higher peak value and it decreased faster compared to the experimental trend. In addition, it can be noted that the oscillations shown in the calculated time trend occurred about 30 seconds earlier and with a greater amplitude than the experimental data. Similar behavior can be observed in **Figure 15**, where the cold leg and the hot leg temperatures in one of the secondary loops are displayed. The cold leg temperature followed the time trends of the DHX sodium outlet temperatures, which had been specified as boundary conditions. The hot leg temperatures decreased quickly at the beginning of the transient reaching the cold leg temperature in about 200 seconds due to the reduction in heat transferred from the primary to the secondary systems across the IHXs, as the primary loop flow rates decreased, and the secondary pumps remained at full speed. It can be noted that, also in this case, the fluctuations of the SM results occur earlier than

A sensitivity simulation to account for the thermal inertia of the temperature instrumentation has been performed to investigate the origin of the discrepancies between the SM results and the experimental data. Beyond the reactor vessel, sodium temperatures were measured in the hot and cold legs of all primary and secondary loops by the Resistance Temperature Detectors (RTDs). The RTDs were spring loaded against the bottom of a thermowell to provide a short response time. Unfortunately, there is no information in the benchmark specifications on the geometry of the temperature detectors, so, after a quick search in the literature and some assumptions, a cylindrical heat structure of 5 cm in diameter was inserted at each RTD location. The sodium temperatures detected by the heat structures are shown in **Figure 14** and **Figure 15** (dotted lines) for the primary loop and secondary loop, respectively. When considering the thermal inertia of the RTDs, the SM

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

well by the simulation.

the measured data.

due to a.

**Figure 13.** *Row2 PIOTA Outlet Temperature.*

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

*Recent Advances in Numerical Simulations*

the active length of all the assemblies.

the DHXs, to set the appropriate sodium temperature.

**3.2 Reference results and sensitivity analysis of the FFTF SM**

Instrumented Open Test Assemblies (PIOTA) and the Row 6 fast response PIOTA. As for the hydraulic part, one heat structure was inserted in each channel to simulate the active part of the assemblies. A flat axial power profile was imposed along

The three primary loops and secondary loops were modeled separately, each one having the same number of hydraulic components. Regarding the secondary loops, two time-dependent components were inserted to provide the proper boundary conditions, one component at the exit of the hot leg piping to set the secondary side pressure, and one component at the beginning of the cold leg piping, downstream

After imposing the boundary conditions (i.e., pumps speed, core power and secondary loops flow conditions) provided by the benchmark team, acceptable steady-state conditions were achieved before performing the transient simulation.

The FFTF LOFWOS Test #13 was initiated when the primary pumps tripped simultaneously. The row 2 PIOTA outlet temperature shown in **Figure 13** can be

The initial rapid rise of the PIOTA outlet temperature was caused by the increasing core power-to-flow ratio following the pump trips. The outlet temperature peaked at around 10 seconds when the power-to-flow ratio reached its maximum value. Then, the increase in core temperatures together with the drop of the GEM sodium level introduced a large negative reactivity feedback, so power decreased faster than the primary flow rate. The drop in reactor power was quick enough to compensate for the reduced flow rate in the primary loop and the sodium temperature started to decrease. As the GEM sodium level approached the bottom of the core, the negative reactivity insertion slowed down. The core outlet temperature began to rise again, and a second peak occurred when the natural circulation was established. Natural circulation was maintained while power continued to decrease resulting in a decrease of the core outlet temperature until

**Figure 13** shows the comparison of the PIOTA outlet temperature predicted by the SM against the measurement (in this and subsequent figures the solid line shows the experimental data while the dashed line displays the SM results). The SM results are in good agreement with the experimental data for the entire duration of

The RELAP5 system thermal hydraulic code was used to make the analysis.

observed to describe the behavior of the FFTF core during the transient.

**254**

**Figure 13.**

*Row2 PIOTA Outlet Temperature.*

the end of the test.

the transient. In particular, the time of occurrence of the two peaks is captured very well by the simulation.

The cold leg and hot leg temperatures in one of the primary loops are shown in **Figure 14**. The hot leg fluid temperature has been quite well predicted by the SM, showing a trend slightly oscillating around the experimental value. That may be due to a.

different prediction of the sodium mixing and thermal stratification phenomena in the outlet plenum of the reactor vessel during the natural circulation phase that are difficult to simulate by a system thermal–hydraulic code. The calculated cold leg fluid temperature showed a faster rise at the beginning of the transient following the increase in the DHX sodium outlet temperatures, it reached a higher peak value and it decreased faster compared to the experimental trend. In addition, it can be noted that the oscillations shown in the calculated time trend occurred about 30 seconds earlier and with a greater amplitude than the experimental data. Similar behavior can be observed in **Figure 15**, where the cold leg and the hot leg temperatures in one of the secondary loops are displayed. The cold leg temperature followed the time trends of the DHX sodium outlet temperatures, which had been specified as boundary conditions. The hot leg temperatures decreased quickly at the beginning of the transient reaching the cold leg temperature in about 200 seconds due to the reduction in heat transferred from the primary to the secondary systems across the IHXs, as the primary loop flow rates decreased, and the secondary pumps remained at full speed. It can be noted that, also in this case, the fluctuations of the SM results occur earlier than the measured data.

A sensitivity simulation to account for the thermal inertia of the temperature instrumentation has been performed to investigate the origin of the discrepancies between the SM results and the experimental data. Beyond the reactor vessel, sodium temperatures were measured in the hot and cold legs of all primary and secondary loops by the Resistance Temperature Detectors (RTDs). The RTDs were spring loaded against the bottom of a thermowell to provide a short response time. Unfortunately, there is no information in the benchmark specifications on the geometry of the temperature detectors, so, after a quick search in the literature and some assumptions, a cylindrical heat structure of 5 cm in diameter was inserted at each RTD location. The sodium temperatures detected by the heat structures are shown in **Figure 14** and **Figure 15** (dotted lines) for the primary loop and secondary loop, respectively. When considering the thermal inertia of the RTDs, the SM

**Figure 14.** *Primary Loop Hot and Cold Leg Temperatures.*

**Figure 15.** *Secondary Loop Hot and Cold Leg Temperatures.*

results are in a very good agreement with the experimental data, improving both the timing and the amplitude of the temperature oscillations.

Then, an additional sensitivity simulation was performed to study the effect of the modeling choice made for the reactor vessel outlet plenum and its impact on sodium mixing. As mentioned before, in the reference case, the reactor vessel and the upper plenum were modeled with a cylindrical multi-dimensional component having three radial meshes. In the two sensitivity simulations, the 3D volumes of the upper plenum were replaced by a 1D vertical pipe component in the first simulation and by a 1D single-volume component in the second simulation. **Figure 16** shows the comparison of the sodium temperature in the hot leg primary loop #1 among the three different reactor vessel outlet plenum modeling choices and with the experimental data. In the reference case (3D) thermal stratification occurs with the hot sodium that tends to go upwards (it should be remembered that the hot leg connection is at the bottom of the outlet plenum, just above the core outlet). In the first sensitivity (1D PIPE), no thermal stratification is observed and the hot sodium exiting the core does not mix with the upper cold sodium at the triggering of natural circulation. In the second sensitivity (1D Single Volume) the hot sodium that would be deposited on top of the reactor completely mixes with the core outlet flow, thus resulting in a slightly higher sodium temperature in the hot leg after the beginning of the transient, compared to the reference case, and which continues to gradually increase even during natural circulation.

**257**

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

**4. Simulation of the CEFR Start-UP Tests with the Serpent Code**

The Neutronics Benchmark of CEFR Start-Up Tests is a CRP proposed by the China Institute of Atomic Energy (CIAE), under the direction and support from IAEA. The CRP was launched in 2018. The main objective of this benchmark is to improve the understanding of the start-up of a SFR and validate the fast reactor analysis computer codes against experimental data obtained at the China

Experimental Fast Reactor (CEFR). The CEFR is the first Chinese fast reactor; it is a pool-type sodium cooled reactor, with a nominal thermal power of 65 MWth [14]. NINE, in collaboration with University of Pisa, participated in all the proposed work packages and, in turn, proposed and organized a work package focused on sensitivity and uncertainty analysis of the first criticality test, that, for the time

The tests included in this benchmark were part of the reactor start-up tests, which included both the fuel loading and the first criticality, the control rod worth measurement, the reactivity coefficients measurement, and the foil activation analysis. All the details of these tests are reported in the Technical specifications [14]. In this chapter, only a sub-set of these tests are analyzed, and their results are

The first test described is the one referred to the "Fuel loading and criticality" (here and after called work-package 1, WP1). It is focused on the analysis of the first criticality achievement; the tests performed are composed by ten sub-critical steps, with different number of fuel Sub-Assemblies (SAs) loaded, and 3 supercritical steps, which have different RE2 control SA insertion levels. The critical RE2 position was found and reported in the Technical Specification through an extrapolation of the experimental super-critical steps. In this work the results for the super-critical steps, as well as the critical ones, are compared with the experimental measurement; a comparison between two different nuclear libraries is also reported, to exploit the dependence on the nuclear data of the effective multiplica-

The second test analyzed is the "Control rod worth measurement", (workpackage 2, WP2). The main goal of this test is to evaluate the control rod worth of each control rod SA and of different group of control rod SAs. Also, in this case a comparison with the experimental measurement is reported in the following

The last test presented is the "Foil activation measurements" (work-package 6, WP6): it concerns the foil activation analysis made through the irradiation of different material samples inside the reactor core in both the axial and radial directions.

The core geometry has been modeled in its full 3D configuration keeping the heterogeneity of most of the present structures (e.g., the hollowed pellet geometry has been modeled). The SAs model starts from the region above the nozzle (the nozzle is not modeled) and reaches the corresponding head. Only few regions of the core, considered less relevant for the simulations, have been homogenized for the sake of simplification (e.g., SAs Handling head, Spring, etc.). The spacer wires have been homogenized with the corresponding cladding, to guarantee the conservation of the stainless-steel mass. **Figure 17(a)** shows the horizontal section of the core taken at a height of 105.1 cm from the bottom; the operation layout is shown, with all the fuel SAs already loaded in the core. **Figure 17(b)** shows a vertical section crossing the center of the core, along the x axis: the homogenized components

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

compared to experimental measurements.

**4.1 The simulation models developed for CEFR**

are noticeable, mostly on the upper part of each assembly.

being, has just begun.

tion factor.

sections.

**Figure 16.** *Sensitivity on Upper Plenum Modeling: Comparison of Primary Loop Hot Leg Temperature.*

*Recent Advances in Numerical Simulations*

*Secondary Loop Hot and Cold Leg Temperatures.*

**Figure 15.**

results are in a very good agreement with the experimental data, improving both

Then, an additional sensitivity simulation was performed to study the effect of the modeling choice made for the reactor vessel outlet plenum and its impact on sodium mixing. As mentioned before, in the reference case, the reactor vessel and the upper plenum were modeled with a cylindrical multi-dimensional component having three radial meshes. In the two sensitivity simulations, the 3D volumes of the upper plenum were replaced by a 1D vertical pipe component in the first simulation and by a 1D single-volume component in the second simulation. **Figure 16** shows the comparison of the sodium temperature in the hot leg primary loop #1 among the three different reactor vessel outlet plenum modeling choices and with the experimental data. In the reference case (3D) thermal stratification occurs with the hot sodium that tends to go upwards (it should be remembered that the hot leg connection is at the bottom of the outlet plenum, just above the core outlet). In the first sensitivity (1D PIPE), no thermal stratification is observed and the hot sodium exiting the core does not mix with the upper cold sodium at the triggering of natural circulation. In the second sensitivity (1D Single Volume) the hot sodium that would be deposited on top of the reactor completely mixes with the core outlet flow, thus resulting in a slightly higher sodium temperature in the hot leg after the beginning of the transient, compared to the reference case, and which continues to gradually increase even during natural circulation.

the timing and the amplitude of the temperature oscillations.

*Sensitivity on Upper Plenum Modeling: Comparison of Primary Loop Hot Leg Temperature.*

**256**

**Figure 16.**
