diff --git a/doc/content/bib/vtb.bib b/doc/content/bib/vtb.bib index 8311e005d..caeff17f4 100755 --- a/doc/content/bib/vtb.bib +++ b/doc/content/bib/vtb.bib @@ -2250,6 +2250,7 @@ @article{Harp2017FCCI publisher={Elsevier} } +<<<<<<< HEAD %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%% % references for Grizzly model of LWR reactor pressure vessel (RPV) @@ -2309,4 +2310,81 @@ @techreport{favpro_v1.0_2024 title = {{FAVPRO} v1.0 User's Manual}, url = {https://www.nrc.gov/docs/ML2411/ML24113A237.pdf}, year = {2024} +======= +%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%% +% references for the EBR-II SHRT + +@article{osti_801571, +title = {Shutdown and closure of the experimental breeder reactor - II.}, +author = {Michelbacher, J A and Baily, C E and Baird, D K and Henslee, S P and Knight, C J and Rosenberg, K E}, +abstractNote = {The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m{sup 3} (86,000 gallons) of sodium and the secondary system contained 50 m{sub 3} (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P and IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.}, +doi = {10.1115/ICONE10-22462}, +url = {https://www.osti.gov/biblio/801571}, +journal = {ICONE-10}, +place = {United States}, +year = {2002}, +month = {9} +} + +@techreport{summer2012benchmark, +title={Benchmark specifications and data requirements for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R}, +author={Sumner, Tyler S and Wei, Thomas YC}, +year={2012}, +institution={Argonne National Lab.(ANL), Argonne, IL (United States)} +} + +@article{mochizuki2014benchmark, + title={Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R}, + author={Mochizuki, Hiroyasu and Muranaka, Kohmei and Asai, Takayuki and Van Rooijen, WFG}, + journal={Nuclear Engineering and Design}, + volume={275}, + pages={312--321}, + year={2014}, + publisher={Elsevier} +} + +@article{mochizuki2018benchmark, + title={Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R--(2) subchannel analysis of instrumented fuel sub-assembly}, + author={Mochizuki, Hiroyasu and Muranaka, Kohmei}, + journal={Nuclear Engineering and Design}, + volume={330}, + pages={14--27}, + year={2018}, + publisher={Elsevier} +} + +@article{mochizuki2010development, + title={Development of the plant dynamics analysis code NETFLOW++}, + author={Mochizuki, Hiroyasu}, + journal={Nuclear Engineering and Design}, + volume={240}, + number={3}, + pages={577--587}, + year={2010}, + publisher={Elsevier} +} + +@techreport{wheeler1976cobra, + title={COBRA-IV-I: An interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores}, + author={Wheeler, CL and Stewart, CW and Cena, RJ and Rowe, DS and Sutey, AM}, + year={1976}, + institution={Battelle Pacific Northwest Labs., Richland, Wash.(USA)} +} + +@article{tano2024validation, + title={Validation of Pronghorn’s subchannel code using EBR-II shutdown heat removal tests: SHRT-17 and SHRT-45R}, + author={Tano, Mauricio and Kyriakopoulos, Vasileios and McCay, James and Arment, Tyrell}, + journal={Nuclear Engineering and Design}, + volume={416}, + pages={112783}, + year={2024}, + publisher={Elsevier} +} + +@techreport{atz2021ducted, + title={Ducted Assembly Steady-State Heat Transfer Software (DASSH): Theory Manual}, + author={Atz, Milos and Smith, Micheal A and Heidet, Florent}, + year={2021}, + institution={Argonne National Lab.(ANL), Argonne, IL (United States)} +>>>>>>> a6bc29d8 (Add documentation for EBR-II model) } diff --git a/doc/content/media/subchannel/EBR-II_primary_tank.png b/doc/content/media/subchannel/EBR-II_primary_tank.png new file mode 100644 index 000000000..4c93010be Binary files /dev/null and b/doc/content/media/subchannel/EBR-II_primary_tank.png differ diff --git a/doc/content/media/subchannel/Normalized_Transients.png b/doc/content/media/subchannel/Normalized_Transients.png new file mode 100644 index 000000000..41b306132 Binary files /dev/null and b/doc/content/media/subchannel/Normalized_Transients.png differ diff --git a/doc/content/media/subchannel/Transient_Temperature.png b/doc/content/media/subchannel/Transient_Temperature.png new file mode 100644 index 000000000..111c4b5a4 Binary files /dev/null and b/doc/content/media/subchannel/Transient_Temperature.png differ diff --git a/doc/content/media/subchannel/Transient_Temperature45.png b/doc/content/media/subchannel/Transient_Temperature45.png new file mode 100644 index 000000000..88c496cbb Binary files /dev/null and b/doc/content/media/subchannel/Transient_Temperature45.png differ diff --git a/doc/content/media/subchannel/XX09.png b/doc/content/media/subchannel/XX09.png new file mode 100644 index 000000000..87f9d6be3 Binary files /dev/null and b/doc/content/media/subchannel/XX09.png differ diff --git a/doc/content/media/subchannel/XX09_3.png b/doc/content/media/subchannel/XX09_3.png new file mode 100644 index 000000000..f9ca1d885 Binary files /dev/null and b/doc/content/media/subchannel/XX09_3.png differ diff --git a/doc/content/media/subchannel/XX09_TTC.png b/doc/content/media/subchannel/XX09_TTC.png new file mode 100644 index 000000000..6e8302589 Binary files /dev/null and b/doc/content/media/subchannel/XX09_TTC.png differ diff --git a/doc/content/media/subchannel/XX09_TTC45.png b/doc/content/media/subchannel/XX09_TTC45.png new file mode 100644 index 000000000..2d35152f2 Binary files /dev/null and b/doc/content/media/subchannel/XX09_TTC45.png differ diff --git a/doc/content/resources/codes_used.md b/doc/content/resources/codes_used.md index d51db0d12..ae5ed3de4 100644 --- a/doc/content/resources/codes_used.md +++ b/doc/content/resources/codes_used.md @@ -99,9 +99,9 @@ obtained through INL's [NCRC](https://inl.gov/ncrc/). - Molten Salt Reactor Experiment RZ multiphysics core model [documentation](msr/msre/multiphysics_rz_model/index.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/msr/msre/multiphysics_core_model/steady_state) -### Pronghorn subchannel +### SCM -- Subchannel ORNL 19 pins and Toshiba 37 pins benchmarks [documentation](sfr/subchannel/index.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/sfr/subchannel) +- SCM validation found in [documentation](sfr/subchannel/index.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/sfr/subchannel) ## Systems analysis and 1D Thermal-hydraulics diff --git a/doc/content/resources/input_features.md b/doc/content/resources/input_features.md index 472a3476f..83f8f9376 100644 --- a/doc/content/resources/input_features.md +++ b/doc/content/resources/input_features.md @@ -172,7 +172,7 @@ specified. The MRAD and PBMR-400 models listed below are an example of this. - 1D TRISO fuel depletion [documentation](htgr/triso/triso_model.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/htgr/triso_fuel) -- Subchannel ORNL 19 pins and Toshiba 37 pins benchmarks [documentation](sfr/subchannel/index.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/sfr/subchannel) +- SCM validation found in [documentation](sfr/subchannel/index.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/sfr/subchannel) - Dispersed UO2 LEU pulse model [documentation](leu_pulse/index.md) and [inputs](https://github.com/idaholab/virtual_test_bed/tree/main/htgr/leu_pulse) diff --git a/doc/content/sfr/subchannel/EBR-II/EBR-2.md b/doc/content/sfr/subchannel/EBR-II/EBR-2.md new file mode 100644 index 000000000..3315a7de6 --- /dev/null +++ b/doc/content/sfr/subchannel/EBR-II/EBR-2.md @@ -0,0 +1,158 @@ + +# EBR-II, SHRT-17 Subchannel model Validation + +*Contact: Vasileios Kyriakopoulos, vasileios.kyriakopoulos.at.inl.gov* + +*Model link: [EBR-II SHRT Subchannel Model](https://github.com/idaholab/virtual_test_bed/tree/devel/sfr/subchannel/EBR-II)* + +!tag name=Subchannel Model for the EBR-II SHRT-17/SHRT-45R + description=Subchannel Model for the EBR-II Shutdown Heat Removal Tests + image=https://mooseframework.inl.gov/virtual_test_bed/media/subchannel/EBR-II_primary_tank.png + pairs=reactor_type:SFR + reactor:EBR-II + geometry:assembly + simulation_type:thermal_hydraulics + transient:steady_state;PLOF;ULOF + V_and_V:validation + codes_used:SCM + computing_needs:Workstation + fiscal_year:2024 + sponsor:NEAMS + institution:INL + +## Test Description + +  + + On April 3, 1986, two tests were carried out to demonstrate the effectiveness of passive feedback in the EBR-II reactor. Both tests began from full reactor power and were initiated when both the primary coolant pumps and the intermediate loop pump were simultaneously tripped, to simulate a loss-of-flow accident. SHRT-17 was a protected loss-of-flow (LOF) test and SHRT-45R was an unprotected loss-of-flow (ULOF) test. At the beginning of test SHRT-17, the primary pumps were tripped at the same time as a full control rod insertion. SHRT-45R, was similar to SHRT-17 except that during this test the plant protection system (PPS) was disabled, to prevent it from initiating a control rod scram. In the second test, reactor power decreased due to reactivity feedback effects. + + The EBR-II reactor vessel grid plenum sub-assembly accommodated 637 hexagonal sub-assemblies, which were installed in one of three regions: a central core, inner blanket, or outer blanket. The central core comprised the 61 sub-assemblies in the first five rows. SCM results are compared with data measured in the XX09 instrumented sub-assembly. More particularly, the code calculations are compared against temperature profile measurements in various axial elevations and the transient temperature evolution of the peak temperature in the central subchannel. + +### Plant Overview + +Argonne National Laboratory’s (ANL) Experimental Breeder Reactor II (EBR-II) was a liquid metal reactor with a sodium-bonded metallic fuel core. EBR-II was rated for a thermal power of 62.5 MW with an electric output of approximately 20 MW. A schematic of the reactor and the primary sodium flow paths are shown in [fig:schematic]. All major primary system components were submerged in the primary tank, which contained approximately $340 m^3$ of liquid sodium at $371^o C$. Two primary pumps inside this pool provided sodium to the two inlet plena of the core. Sub-assemblies in the inner core received sodium from the high-pressure inlet plenum, accounting for approximately $85\%$ of the total primary flow. The blanket and reflector sub-assemblies in the outer blanket region received sodium from the low-pressure inlet plenum. Hot sodium exited the sub-assemblies into a common upper plenum, where it mixed before passing into the intermediate heat exchanger (IHX). + +!media subchannel/EBR-II_primary_tank.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:schematic + caption=Schematic of the EBR-II reactor [!cite](osti_801571) + +The reactor-vessel grid-plenum sub-assembly accommodated 637 hexagonal sub-assemblies. The sub-assemblies were divided into three regions: core, inner blanket (IB) and outer blanket (OB). The central core comprised the 61 sub-assemblies in the first five rows. Two positions in row 3 contained safety-rod sub-assemblies and eight positions in Row 5 contained control-rod sub-assemblies. Two positions in Row 5 contained the instrumented sub-assemblies (INSAT) XX09 and XX10, and one position in Row 5 contained the in-core instrument test facility (INCOT) XY16. The remainder of the central core region contained driver fuel or experimental-irradiation sub-assemblies. EBR-II was heavily instrumented to measure mass flow rates, temperatures, and pressures, throughout the system. + +### Instrumented sub-assemblies + +The SHRT-17 test is a protected LOF test. This test was initiated by a trip of the primary and intermediate pumps under the rated-power of 57.3MW. The reactor was scrammed at the same time as the pump trips. As flow dissipated in the primary system after the pump trips, cooling of the core transitions from forced to natural circulation, while temperature and flow rate converge to an equilibrium state. + +The SHRT-45R test is an unprotected LOF test. Similarly to SHRT-17, this test was initiated by a trip of the primary and intermediate pumps under the rated-power of 60.0 MW, but without scram. In the SHRT-45R experiment, the auxiliary electromagnetic pump (EMP) was kept operational throughout the test duration. As a result, the cooling situation is not fully natural convection, but since the EMP flow rate is very low, it is also not fully forced circulation. Because of the operation of the EMP in the SHRT-45R experiment, the coolant mass flow rate converges to about two times that in the SHRT-17 experiment. + +Both tests run for $900$ seconds. + +This work utilizes data measured by the instrumented sub-assembly XX09. XX09 was a fueled sub-assembly specifically designed with a variety of instrumentation to provide data for benchmark validation purposes. The standard-type fueled sub-assembly contains 91 fuel pins, whereas in XX09 the outer row of fuel pins was removed and a guide thimble was inserted instead, as shown in [fig:XX09], [fig:XX09_3]. + +!media subchannel/XX09.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:XX09 + caption=XX09 Instrumented sub-assembly axial-section [!cite](summer2012benchmark) + +!media subchannel/XX09_3.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:XX09_3 + caption=XX09 Instrumented sub-assembly cross-section [!cite](summer2012benchmark) + +### Boundary conditions + +The values for the inlet mass flow rate, power, and inlet coolant temperature are specified in the EBR-II SHRT benchmark [!cite](summer2012benchmark). The value of outlet pressure has been approximated based on the operation conditions of the EBR-II before the transients. These values represent the steady state conditions leading to the transients. + +| Experimental steady-state parameters | +| | + +| Experiment Parameter (Unit) | SHRT-17 | SHRT-45R | +| - | - | - | +| Inlet Mass flow rate of XX09 (kg/s) | 2.45 | 2.427 | +| Power of XX09 (kW) | 486.2 | 379.8 | +| Inlet coolant temperature (K) | 624.7 | 616.4 | +| Outlet Pressure (kPa) | 200.0 | 200.0 | + +During the SHRT transients, the mass flow rate and power vary. The normalized power and mass flow rate during the transients have been adapted from [!cite](mochizuki2014benchmark) , [!cite](mochizuki2018benchmark) and are presented in [fig:Norm_TR]. This information is used as input for the SCM transient calculations. The work in the cited sources utilizes a NETFLOW++ [!cite](mochizuki2010development) simulation to inform a COBRA-IV-I [!cite](wheeler1976cobra) model of the instrumented sub-assembly XX09. It should be noted that for SHRT-17, the power generation throughout the transient is solely due to decay heat because the protection system shuts down the reactor. For the SHRT-45R test, fission power continues into the transient for some time until the reactivity feedback mechanisms ultimately shut down the reactor. + +!media subchannel/Normalized_Transients.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:Norm_TR + caption=Transient boundary conditions + + +## Steady State Results + +Three simulation results are presented here and compared with experimental measurements: + +- item Results from the [DASSH](https://github.com/dassh-dev/examples/tree/master/Example-3) [!cite](atz2021ducted),subchannel code, which is used as a reference for subchannel agreement. +- item Results from SCM with a uniform (axially and radially) pin power profile. +- Results from SCM only with the corrected power profile (axially and radially) computed via the Serpent-2 model. + +Aditional results can be found in the relevant publication [!cite](tano2024validation) + +The DASSH subchannel code models the internal pin region of sub-assembly XX09 and the thimble region around it. The standalone SCM simulation only models the internal pin region. Both codes don't consider the neighboring sub-assemblies. [fig:TTC17], [fig:TTC45] present the results for the steady-state radial temperature profiles calculated at position TTC, which is near the outlet of the fueled section, for tests SHRT-17 and SHRT-45R, respectively. The figures include calculations obtained from the DASSH subchannel code. DASSH solves for the conservation of axial momentum and energy and assumes that the cross flows are produced to close the mass balance, i.e., there is no explicit equation solved for the conservation of lateral momentum in the code, and advective radial heat transfer due to cross flows is ignored. Radial thermal mixing is modeled by being lumped into the conduction term and approximated by enhancing the thermal diffusivity with an eddy diffusivity obtained from correlation [!cite](atz2021ducted). This approximation is generally accurate for liquid-metal-cooled reactors, and hence DASSH results are used as a baseline for the performance of the subchannel code. + +Expert input from E. Feldman [DASSH](https://github.com/dassh-dev/examples/tree/master/Example-3) suggests that the coolant flow rate in the thimble region is roughly 8-10\% of the sub-assembly total flow rate. Based on this, the sub-assembly total flow rate of the DASH model, is increased from the nominal values to $2.6923kg/s$ for test SHRT-17. This means that the thimble flow rate is artificially rerouted through the interior of sub-assembly XX09. However, the thimble flow is significantly colder than the flow through the XX09 assembly. Thus, the approximation of adding the thimble flow to the sub-assembly, which assumes that the thimble and sub-assembly flow are in thermal equilibrium, might not be accurate. This rerouting was not done for the SCM simulations. + +For SHRT-17, in the constant pin power case, both SCM and DASSH exhibit similar behavior. Since DASSH does not resolve the crossflows (contrary to SCM), similar results indicate that crossflows might not be instrumental in determining the temperature profile for this case. Additionally, DASSH predicts a slightly less skewed distribution than SCM, which is closer to the experimental results. This means that the crossflows may be underestimated by the lateral momentum balance equation solved by SCM, or that the thimble model incorporated in DASSH is more accurate than the simplified, mass-flow adaptation applied to SCM. Nonetheless, both the SCM and DASSH calculations, are close enough to suggest that those differences in modeling approach, do not produce large discrepancies in the results. + +!media subchannel/XX09_TTC.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:TTC17 + caption=Test SHRT-17 + +!media subchannel/XX09_TTC45.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:TTC45 + caption=Test SHRT-45R + +The above SCM results can be further improved by coupling to a Pronghorn model. This model can calculate the heat flux from the edge subchannels to the inner duct of the thimble, and apply this to the SCM simulations. This coupled simulation improves the SCM calculation specifically for the subchannels closest to the duct [!cite](tano2024validation). + +## Transient Results + +Simulations a uniform (radial and axial) pin power profile are performed with the SCM model. The results are compared against experimental measurements and the NETFLOW++/COBRA simulations which are used as a reference result. The temperature transient calculations, together with the experimental measurements at thermocouple location TTC-31, are presented in [fig:transient1],[fig:transient2]. In SHRT-17, the power of the XX09 sub-assembly decreases rapidly at time zero due to the reactor scram, which results in a sudden coolant temperature decrease. On the other hand, in SHRT-45R there is no scram of the reactor, so power decreases at a much lower pace. This is why no sudden temperature drop is observed at time zero. + +Then for both tests, the flow rate gradually decreases due to the pump trip and coasts-down, causing the coolant temperature to increase to a peak of around $890\mathrm{K}$ for SHRT-17 and $970\mathrm{K}$ for SHRT-45R. Following that, the coolant temperature decreases and levels off to $700\mathrm{K}$ for SHRT-17 and $730\mathrm{K}$ for SHRT-45R with the establishment of natural circulation due to decay heat and buoyancy effects. + +For SHRT-17 and SHRT-45R, the stand-alone SCM transient, underestimates the measured result of the peak temperature. For the SHRT-17 transient, the uniform power model under-predicts the peak temperature by approximately $3\mathrm{K}$. For the SHRT-45 transient, the uniform power model under-predicts the peak temperature by approximately $30\mathrm{K}$. In general subchannel temperatures are expected to be lower than the measurements since SCM calculates surface averaged temperatures and the experiment took point-wise measurements on thermocouples attached to the heated pins. + +The thermal inertia of the solid structures play an important role in determining the temperature field transient, towards the establishment of natural convection. As these structures are not modeled in the SCM simulations, a more rapid decrease in the temperature field is observed in all cases. Nonetheless, the thermal inertia of these structures does not play a fundamental role once natural convection is established, as the system is in thermal equilibrium. Hence, better agreement is obtained for this part of the transient. + +In general, good results are obtained for both transients. + +!media subchannel/Transient_Temperature.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:transient1 + caption=Test SHRT-17 + +!media subchannel/Transient_Temperature45.png + style=width:60%;margin-bottom:2%;margin:auto; + id=fig:transient2 + caption=Test SHRT-45R + +As previously mentioned, the SHRT-45R test is an unprotected loss of flow test, meaning that there is no reactor SCRAM. Therefore, we observe higher temperatures compared to the SHRT-17 test in which SCRAM is performed. The power transient used in the SCM calculation is adapted from [!cite](mochizuki2014benchmark), [!cite](mochizuki2018benchmark). Additionally, during the SHRT-45R transient, the EM pump is kept on with low power to maintain a slightly higher flow rate than natural circulation conditions. Initially, the coolant flow rate decreases, causing the temperature in the core to rise. The temperature increase leads to negative feedback due to the thermal expansion of the fuel sub-assembly, decrease of coolant density, and the Doppler effect in the nuclear cross sections. At approximately 60 seconds into the transient, the negative feedback effect becomes strong enough to significantly reduce the core power, and the temperature begins to decrease. Around 650 seconds into the SHRT-45R test, there is a sudden dip in the temperature because the auxiliary EM pump power is increased, and the coolant mass flow rate increases rapidly. + +For the SHRT-17 test, both the primary coolant and intermediate-loop pumps trip, simulating a loss-of-flow accident. Additionally, the primary system auxiliary pump is deactivated, and the SCRAM signal is sent to the reactor to reduce power. The temperature increases rapidly due to the fast reduction in flow rate. As the reactor power reduces and natural convection is established, the temperature decreases toward a new steady-state value. A peak in the temperature is reached approximately 40 seconds after the start of the transient. + +## Input files + +To run the steady state problem use the following input files: + +!listing sfr/subchannel/EBR-II/XX09_SCM_SS17.i language=cpp + +!listing sfr/subchannel/EBR-II/XX09_SCM_SS45R.i language=cpp + +To run the transient problem use the following input files: + +!listing sfr/subchannel/EBR-II/XX09_SCM_TR17.i language=cpp + +!listing sfr/subchannel/EBR-II/XX09_SCM_TR45R.i language=cpp + +The corrected radial power profile is read by the following .txt file (pin_power_profile61.txt): + +!listing sfr/subchannel/EBR-II/pin_power_profile61.txt language=cpp + +The uniform radial power profile is read by the following .txt file (pin_power_profile61_uniform.txt): + +!listing sfr/subchannel/EBR-II/pin_power_profile61_uniform.txt language=cpp \ No newline at end of file diff --git a/doc/content/sfr/subchannel/index.md b/doc/content/sfr/subchannel/index.md index eafc31531..9f28d21d2 100644 --- a/doc/content/sfr/subchannel/index.md +++ b/doc/content/sfr/subchannel/index.md @@ -4,4 +4,4 @@ [Toshiba 37-pin benchmark](toshiba_37_pin/toshiba_37_pin.md) -[ORNL Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility blockage, benchmark](thors/thors.md) +[EBR-II shutdown heat removal tests (SHRT-17/SHRT-45R)](EBR-II/EBR-2.md) diff --git a/doc/content/sfr/subchannel/thors/thors.md b/doc/content/sfr/subchannel/thors/thors.md index 6f5939658..b6a407530 100644 --- a/doc/content/sfr/subchannel/thors/thors.md +++ b/doc/content/sfr/subchannel/thors/thors.md @@ -2,7 +2,7 @@ *Contact: Vasileios Kyriakopoulos, vasileios.kyriakopoulos.at.inl.gov* -*Model link: [THORS edge blockage Subchannel Model](https://github.com/idaholab/virtual_test_bed/tree/devel/sfr/subchannel/THORS)* +*Model link: [THORS edge blockage Subchannel Model](https://github.com/idaholab/virtual_test_bed/tree/devel/sfr/subchannel/thors)* !tag name=Effect of Partial Blockages in Simulated LMFBR Fuel Assemblies description=Study of the partial blockage in sodium fast reactor assemblies using a subchannel discretization of the thermal hydraulics. The new flow distribution is computed and analyzed