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Fuel cladding damage assessment based on reactivity insertion accidents for a 10 MWth LBE cooled fast reactor
Annals of Nuclear Energy 129 (2019) 412–417
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Fuel cladding damage assessment based on reactivity insertion accidents for a 10 MWth LBE cooled fast reactor Ting Li a,b, Guowei Wu a, Chao Liu a, Jin Wang a,⇑ a b
Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031, China University of Science and Technology of China, Hefei, Anhui 230027, China
a r t i c l e
i n f o
Article history: Received 5 September 2018 Received in revised form 25 January 2019 Accepted 30 January 2019 Available online 18 February 2019 Keywords: LBE cooled fast reactor Fuel cladding damage Accident assessment RIA RELAP5
a b s t r a c t Fuel cladding integrity affects the safety and economy of nuclear reactors, especially in accident condition. In this paper, based on the reactivity insertion accident (RIA) analysis, the fuel cladding damage has been analyzed for a 10 MWth LBE cooled fast reactor developed by Institute of Nuclear Energy Safety and Technology (INEST) using RELAP5. Simulation was carried out for control rod withdrawal accident with a reactivity insertion rate of 50.19 pcm/s for 4 s, both for scram and without scram. During the control rod withdrawal accident without scram, the highest cladding temperature was 714 °C, which was below the temperature limit of cladding integrity. It is concluded that the RIA will not cause reactor safety violation due to the reactor’s inherent safety feature. Ó 2019 Elsevier Ltd. All rights reserved.
1. Introduction The lead-bismuth eutectic (LBE) has obvious advantages of excellent thermal-physical, chemical properties and forgiving core behavior mainly owing to its negative reactivity coefficient (IAEA, 2012; OECD/NEA, 2007; Zhang, 2013). As a result, Lead cooled Fast Reactor (LFR) was selected as one of the six Generation-IV reactor concept candidates (OECD/NEA, 2014). A Strategic Priority Research Program called the ‘‘Future Advanced Nuclear Fission Energy—ADS Transmutation System” was initiated by Chinese Academy of Sciences (CAS) (Wu et al., 2014, 2015a,b) in 2011, aiming to develop and master all the key technologies of the ADS by three phases of research and development. The China LEAd-based Reactor (CLEAR), proposed and developed by Institute of Nuclear Energy Safety Technology (INEST)FDS Team, CAS, was chosen as the reference reactor system of the ADS project (Huang et al., 2013; Wu and FDS Team, 2009; Wu et al., 2016; Wu, 2016). In the first phase, a 10 MWth LBE cooled research reactor was designed for experiments on neutronics, thermal hydraulics and safety characteristics. Reactivity insertion accident (RIA) is a nuclear reactor accident that involves an unwanted increase in fission rate and reactor power. Control rod withdrawal accident is one of the main accident scenarios for RIA (OECD/NEA, 2010). Many accident analyses for LBE reactors have been conducted based on RIAs to study the safety performance of reactor’s fuel cladding. A spurious with⇑ Corresponding author. E-mail address: [email protected] (J. Wang). https://doi.org/10.1016/j.anucene.2019.01.056 0306-4549/Ó 2019 Elsevier Ltd. All rights reserved.
drawal of the most reactive control rod was analyzed for 300 MWth lead cooled European advanced demonstrator reactor to find out its safety feature (European SEVENTH FRAMEWORK PROGRAMME, 2013). The RIA was studied for an 800 MWth ADS subcritical LBE cooled reactor to investigate the reactor safety (Li et al., 2015; Lu et al., 2016). The unprotected overpower transient, caused by withdrawal of one control rod, was performed by a Neutronic and Thermal-hydraulics software called NTC-2D to investigate the safety behaviors of the natural circulation LBE reactor, with a coolant flow rate change of 38.9% (Gu et al., 2015). However, for a small forced circulation reactor, the coolant mass flow characteristics regulated by pumps are different from natural circulation reactor in accident condition, resulting in different behavior of reactor fuel cladding temperature changes. As a result, the RIA is analyzed by RELAP5 for the 10 MWth LBE cooled reactor. Two kinds of RIA named as control rod withdrawal accident with scram and that without scram, are simulated, with a reactivity insertion rate of 50.19 pcm/s. The accident without scram has great difference from that with scram in fuel cladding peak temperature and reactor power. By comparing the two accidents results, the reactor inherent safety feature will be investigated. 2. Calculation model 2.1. Reactor design The main components of the LBE cooled reactor are showed in Fig. 1. It mainly consists of the core, the LBE coolant, primary heat
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exchangers (PHXs) and the Reactor Vessel Air Cooling System (RVACS). RVACS is designed for removing decay heat. It includes an air riser tube, a thermal insulating layer, an air down-comer tube and a chimney. The core fuel assembly (FA) layout (Fig. 2) is designed by VisualBUS (Wu et al., 2007), with a minimized fuel inventory and adjusted power distribution. The core consists of a spallation target, fuel assemblies, reflection assemblies, and shielding assemblies. The FA configuration is shown in Fig. 3. It includes 61 fuel pins, with UO2 pellets as the fuel and the 15-15Ti stainless steel as the cladding. The preliminary structure of FA is made up of the following six parts: an operation head, a top nozzle, a fuel element, a bottom nozzle, a shroud and a bottom spike moving from the top to the bottom. The fuel element consists of eight components in a stainless steel cladding tube, namely the upper and lower end caps, a tighten spring, fuel pellets, upper and lower reflectors, a gas chamber and a counter weight. To ensure the inherent reactivity control, the fuel doppler coefficient and coolant temperature coefficient are designed to be negative. The reactor main design parameters are shown in Table 1, and the neutron parameters are shown in Table 2 (Wang et al., 2016; Wu et al., 2016, 2017). The maximum power of reactor fuel pin is 2.65 kW and that of the assembly is 159.25 kW.
Fig. 2. Core fuel assembly layout of the reactor.
2.2. RELAP5 model A detailed RELEAP5 model of the reactor (Fig. 4) (Wu et al., 2015a,b) was established for the LBE cooled research reactor using RELAP5/MOD 4.0 (Chen et al., 2014). In this model, the core model was simplified with the hottest pin, the hottest assembly and average assembly, adding core power to structure numbered 901/902/903 to simulate core. Four primary heat exchangers (HX1 HX4) were modeled to stand for the PHXs. The heat exchange part of each PHX was divided into 10 sections axially. The secondary side inlet of the heat exchanger was set to a constant temperature and flow value. The outlet side was set to a constant pressure. The secondary loops boundary conditions were represented with time dependent volume (TDV) and timeFig. 3. Configuration of fuel assembly.
Table 1 LBE cooled reactor design parameters.
Fig. 1. LBE-cooled reactor schematic diagram.
Parameters
Value
Unit
Power Core inlet/outlet temperature Primary coolant mass flow Secondary inlet/outlet temperature Secondary loop coolant pressure Secondary loop coolant mass flow Control rods falling time Fuel Number of fuel assemblies Number of primary pumps Primary coolant Secondary coolant
10 300/385 811.67 215/230 4 170 10 UO2 86 2 LBE Pressurized liquid water
MW °C kg/s °C MPa kg/s s – – – – –
dependent junctions (TDJ). The primary pumps (107/117) boundary conditions were represented with TDJ. The nodalization diagram of the RELAP5 model of RVACS is shown in Fig. 5. The air supply and exhaust were represented with two TDVs, components 610 and 650, which were set to the atmospheric pressure. RVACS inlet temperature was set to 50 °C. The RVACS air riser and down-comer tubes were represented by components 620 and 630, respectively.
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Table 2 LBE cooled reactor neutron parameters. Parameters
Value
Unit
Effective delayed neutron fraction Doppler constant Coolant expansion coefficient Axial core expansion coefficient Radial core expansion coefficient
706 0.44 0.53 0.072 0.788
pcm pcm/°C pcm/°C pcm/°C pcm/°C
Scram signal is the power value that increases up to 1.1 times than the normal operation power value (11 MW), and the core outlet temperature that exceeds 395 °C. It is assumed that there is 1.2 s delay for signal transfer of scram command after a trigger for the same is generated. Fuel damage should be found out by studying peak cladding temperature in the core in the process of accident evolution. The standard for fuel cladding damage (referring to European SEVENTH FRAMEWORK PROGRAMME, 2013) is that the cladding temperature must not exceed 800 °C. 3. Transient thermal-hydraulic analysis 3.1. Steady state simulation To study the calculation foundation for accident analysis, the steady state simulation of the LBE reactor was performed at the initial fuel life using RELAP5. The reactor design and simulation values comparisons are shown in Table 3. It shows that the simulation values were in good agreement with design values.
Fig. 5. Nodalization diagram of RELAP5 model of RVACS.
3.2. Accident analysis 3.2.1. Control rod withdrawal accident with scram In this paper, the RIA was studied that some control rods were withdrawn causing the increase of core reactivity. According the reactor design, total reactivity should be less than 0.002 4K/K in
reactivity insertion accident. With regard to the accident, it was assumed that the reactivity was inserted linearly, with a rate of 200.76 pcm in 4 s. The reactor was shut down during the transient accident according to the scram triggered signal.
Fig. 4. Nodalization diagram of RELAP5 model of the reactor.
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Table 3 Comparison of simulated values against reactor design parameters. Parameters
Design Value
RELAP5
Difference (%)
Power (MW) Average inlet temperature (°C) Average outlet temperature (°C)
10 300 385
9.70 301 384
3 0.33 0.33
The core power and reactivity changes are shown in Fig. 6. In the beginning, the core power rose fast due to the reactivity insertion. When the core power increased up to 11 MW, the scram signal was triggered. After reactor shut down, the secondary systems stopped following the feed water pumps close. Simultaneously, the decay heat was removed by opening the triggering valve of RVACS system. After 1.2 s delay, the reactor scrammed at 2.2 s. Then the core power decreased rapidly, so did the reactor reactivity. The core power climbed to the maximum value of 12.3 MW at about 2.2 s. The reactivity of the reactor held the maximum value of 108.9 pcm. The coolant inlet, coolant outlet, cladding and fuel temperature changes in first 20 s are presented in Fig. 7. The coolant outlet, cladding and fuel maximum temperature reached a maximum value of about 385 °C, 418 °C and 463 °C, respectively. The cladding peak temperatures stabilized far below 800 °C. The RVACS power surpassed the core decay power at about 2.8 s. After that time the cladding temperature began to decline. The coolant mass flow rate is shown in Fig. 8. With the core power increasing, the temperature difference between coolant outlet and inlet became lager leading to coolant buoyancy effect capacity enhancement. As a result, the coolant mass flow rate rose to remove more heat produced by the core in the first 2.2 s, and then decreased for the reactor shutdown and temperature decline, reducing to 803.90 kg/s at the minimum, with a change of 0.979%. And then it rose and tried to return back to the initial flow rate for the mass regulation of pump represented with TDJ in RELAP5 model.
3.2.2. Control rod withdrawal accident without scram The control rod withdrawal accident without scram has the same reactivity insertion rate as that accident with scram. The reactor did not shut down during this accident by scram system. The core power and reactivity changes are shown in Fig. 9. Fig. 10 presents an enlargement of Fig. 9.The reactor reactivity increased fast during the first 4 s because of the reactivity insertion and then decreased due to the doppler effect, coolant and core expansion negative temperature feedback effect (as shown in Fig. 11). Due to above negative feedback effect, the reactivity
Fig. 6. The core power and reactivity changes in control rod withdrawal accident with scram.
Fig. 7. The core inlet, outlet, fuel, and cladding temperature changes in control rod withdrawal accident with scram.
Fig. 8. The coolant mass flow rate change in control rod withdrawal accident with scram.
Fig. 9. The core power and reactivity change in control rod withdrawal accident without scram.
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Fig. 10. An enlargement of Fig. 9. Fig. 12. The core inlet, outlet, fuel, and cladding temperature change in control rod withdrawal accident without scram.
Fig. 11. Reactivity feedback in control rod withdrawal accident without scram.
climbed to the peak value of 193 pcm rather than 200.76 pcm. After 4 s, the reactor total reactivity decreased rapidly due to the negative feedback effect, with slow increase of the reactor power. The maximum value of the core power was 35.8 MW at 250 s. As time went on, the total reactivity decreased to be negative (less than 0 pcm) at 250 s, causing the core power to go down after then. The coolant inlet, coolant outlet, cladding and fuel temperature change in the short time are presented in Fig. 12. The coolant outlet, cladding and fuel maximum temperature reached a maximum value of about 613 °C, 714 °C and 907 °C, respectively, which were all below the safety design limits. The coolant mass flow rate change is shown in Fig. 13. With the core power increasing, the temperature difference between coolant outlet and inlet becomes larger leading to coolant buoyancy effect capacity enhancement. As a result, the coolant mass flow rate rose to remove more heat produced by the core, increasing to 830.42 kg/s at the maximum, with a flow rate change of 2.31%. And then it tried to decrease to the initial flow rate for the mass regulation of pump represented with TDJ in RELAP5 model. 4. Conclusions A detailed RELAP5 model has been established for accident analysis of the 10 MWth LBE cooled reactor in order to evaluate reactor cladding damage. The model is capable of analyzing a wide
Fig. 13. The coolant mass flow rate change in control rod withdrawal accident without scram.
range of accident scenarios. Two related RIAs, named as control rod withdrawal accident with scram and that without scram, have been simulated. Reactor power and peak cladding temperature have been examined in detail. The reactor power grows exponentially, owing to the constant positive reactivity insertion rate of 50.19 pcm per second, and drops very rapidly after reactor scram. The change in the reactor coolant flow rate regulated by pumps is very small with a maximum change of 2.31%, which is far less than that of the natural circulation reactor. Cladding temperature increases due to the energy deposition during the power rise. The maximum cladding temperature during the control rod withdrawal accident without scram is 714 °C, which is below temperature limit. As a result, there is no damage to fuel cladding in both the control rod withdrawal accident with scram and without scram, which will not lead to the leakage of the radionuclide. Therefore, it is concluded that the RIA will not result in safety violation. By comparison with the two accident analysis results, no fuel cladding damage occurs and inherent safety feature of the 10 MWth LBE cooled reactor is proved owing to its negative feedback.
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Acknowledgment This work was supported by the National Key Research and Development Program of China：Megawatt Ultra-small Liquid Metal Cooled Space Nuclear Reactor (Grant No. 2018YFB1900600). We further thank the great help from other members of FDS Team in this research. Appendix A. Supplementary data Supplementary data to this article can be found online at https://doi.org/10.1016/j.anucene.2019.01.056. References Chen, S. et al., 2014. Analysis of neutronics and dynamic characteristics with reactivity injection in LBE cooled sub-critical reactor. Chin. J. Nucl. Saf. 13, 45– 49. European SEVENTH FRAMEWORK PROGRAMME, 2013. REPORT ON THE RESULTS OF ANALYSIS OF DBC EVENTS FOR THE ETDR (ALFRED), Grant agreement no. FP7-249668, 77–80. Gu, Z. et al., 2015. Transient analysis o loss of heat sink and overpower transient of natural circulation LBE-cooled reactor. Prog. Nucl. Energy 81, 60–65. Huang, Q. et al., 2013. Recent progress of R&D activities on reduced activation ferritic/martensitic steels. J. Nucl. Mater. 442, S2–S8. IAEA, 2012. Liquid Metal Coolants for Fast Reactor Cooled by Sodium, Lead and Lead-Bismuth Eutectic, No. NP-T-1.6. IAEA, Vienna.
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Li, X. et al., 2015. Preliminary studies of a new accelerator-driven minor actinide bumer in industrial scale. Nucl. Eng. Des. 292, 57–68. Lu, T. et al., 2016. Preliminary safety analysis on loss of flow accidents and external source transients for LBE cooled ADSR core. Prog. Nucl. Energy 88, 134–146. OECD/NEA, 2007. Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Material Compatibility Thermal Hydraulics and Technologies, NEA No.6195. OECD/NEA, 2010. Nuclear Fuel Behaviour Under Reactivity-initiated Accident (RIA) Conditions, NEA No. 6847. OECD/NEA, 2014. Technology Roadmap Update for Generation IV Nuclear Energy Systems. IAEA, Vienna. Wang, M. et al., 2016. Preliminary conceptual design of a lead-bismuth cooled small reactor (CLEAR-SR). Int. J. Hydrogen Energy 40, 15132–15136. Wu, G. et al., 2017. Primary pump coast-down characteristics analysis in lead cooled fast reactor under loss of flow transient. Ann. Nucl. Energy 103, 1–9. Wu, G. et al., 2015a. Assessment of RVACS performance for small size lead-cooled fast reactor. Ann. Nucl. Energy 77, 310–317. Wu, Y., 2016. Design and R&D progress of china lead-based reactor for ADS research facility. Engineering 2, 124–131. Wu, Y. et al., 2014. Design and R&D progress of China lead-based reactor. 22nd International Conference on Nuclear Engineering. Amer. Soc. Mech. Eng.. V003T06A051. Wu, Y. et al., 2015b. CAD-based Monte Carlo program for integrated simulation of nuclear system SuperMC. Ann. Nucl. Energy 82, 161–168. Wu, Y. et al., 2016. Development strategy and conceptual design of China lead based research reactor. Ann. Nucl. Energy 87, 511–516. Wu, Y.FDS Team, 2009. Fusion-based hydrogen production reactor and its material selection. J. Nucl. Mater. 386, 122–126. Wu, Y., Li, J., Li, Y., et al., 2007. An integrated multifunctional neutronics calculation and analysis code system: VisusalBUS. Chin. J. Nucl. Sci. Eng. 27 (4), 365–373. Zhang, J., 2013. Lead-bismuth eutectic (LBE): a coolant candidate for Gen. IV advanced nuclear reactor concepts. Adv. Eng. Mater. 16, 349–356.
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Fuel cladding damage assessment based on reactivity insertion accidents for a 10 MWth LBE cooled fast reactor Ting Li a,b, Guowei Wu a, Chao Liu a, Jin Wang a,⇑ a b
Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031, China University of Science and Technology of China, Hefei, Anhui 230027, China
a r t i c l e
i n f o
Article history: Received 5 September 2018 Received in revised form 25 January 2019 Accepted 30 January 2019 Available online 18 February 2019 Keywords: LBE cooled fast reactor Fuel cladding damage Accident assessment RIA RELAP5
a b s t r a c t Fuel cladding integrity affects the safety and economy of nuclear reactors, especially in accident condition. In this paper, based on the reactivity insertion accident (RIA) analysis, the fuel cladding damage has been analyzed for a 10 MWth LBE cooled fast reactor developed by Institute of Nuclear Energy Safety and Technology (INEST) using RELAP5. Simulation was carried out for control rod withdrawal accident with a reactivity insertion rate of 50.19 pcm/s for 4 s, both for scram and without scram. During the control rod withdrawal accident without scram, the highest cladding temperature was 714 °C, which was below the temperature limit of cladding integrity. It is concluded that the RIA will not cause reactor safety violation due to the reactor’s inherent safety feature. Ó 2019 Elsevier Ltd. All rights reserved.
1. Introduction The lead-bismuth eutectic (LBE) has obvious advantages of excellent thermal-physical, chemical properties and forgiving core behavior mainly owing to its negative reactivity coefficient (IAEA, 2012; OECD/NEA, 2007; Zhang, 2013). As a result, Lead cooled Fast Reactor (LFR) was selected as one of the six Generation-IV reactor concept candidates (OECD/NEA, 2014). A Strategic Priority Research Program called the ‘‘Future Advanced Nuclear Fission Energy—ADS Transmutation System” was initiated by Chinese Academy of Sciences (CAS) (Wu et al., 2014, 2015a,b) in 2011, aiming to develop and master all the key technologies of the ADS by three phases of research and development. The China LEAd-based Reactor (CLEAR), proposed and developed by Institute of Nuclear Energy Safety Technology (INEST)FDS Team, CAS, was chosen as the reference reactor system of the ADS project (Huang et al., 2013; Wu and FDS Team, 2009; Wu et al., 2016; Wu, 2016). In the first phase, a 10 MWth LBE cooled research reactor was designed for experiments on neutronics, thermal hydraulics and safety characteristics. Reactivity insertion accident (RIA) is a nuclear reactor accident that involves an unwanted increase in fission rate and reactor power. Control rod withdrawal accident is one of the main accident scenarios for RIA (OECD/NEA, 2010). Many accident analyses for LBE reactors have been conducted based on RIAs to study the safety performance of reactor’s fuel cladding. A spurious with⇑ Corresponding author. E-mail address: [email protected] (J. Wang). https://doi.org/10.1016/j.anucene.2019.01.056 0306-4549/Ó 2019 Elsevier Ltd. All rights reserved.
drawal of the most reactive control rod was analyzed for 300 MWth lead cooled European advanced demonstrator reactor to find out its safety feature (European SEVENTH FRAMEWORK PROGRAMME, 2013). The RIA was studied for an 800 MWth ADS subcritical LBE cooled reactor to investigate the reactor safety (Li et al., 2015; Lu et al., 2016). The unprotected overpower transient, caused by withdrawal of one control rod, was performed by a Neutronic and Thermal-hydraulics software called NTC-2D to investigate the safety behaviors of the natural circulation LBE reactor, with a coolant flow rate change of 38.9% (Gu et al., 2015). However, for a small forced circulation reactor, the coolant mass flow characteristics regulated by pumps are different from natural circulation reactor in accident condition, resulting in different behavior of reactor fuel cladding temperature changes. As a result, the RIA is analyzed by RELAP5 for the 10 MWth LBE cooled reactor. Two kinds of RIA named as control rod withdrawal accident with scram and that without scram, are simulated, with a reactivity insertion rate of 50.19 pcm/s. The accident without scram has great difference from that with scram in fuel cladding peak temperature and reactor power. By comparing the two accidents results, the reactor inherent safety feature will be investigated. 2. Calculation model 2.1. Reactor design The main components of the LBE cooled reactor are showed in Fig. 1. It mainly consists of the core, the LBE coolant, primary heat
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exchangers (PHXs) and the Reactor Vessel Air Cooling System (RVACS). RVACS is designed for removing decay heat. It includes an air riser tube, a thermal insulating layer, an air down-comer tube and a chimney. The core fuel assembly (FA) layout (Fig. 2) is designed by VisualBUS (Wu et al., 2007), with a minimized fuel inventory and adjusted power distribution. The core consists of a spallation target, fuel assemblies, reflection assemblies, and shielding assemblies. The FA configuration is shown in Fig. 3. It includes 61 fuel pins, with UO2 pellets as the fuel and the 15-15Ti stainless steel as the cladding. The preliminary structure of FA is made up of the following six parts: an operation head, a top nozzle, a fuel element, a bottom nozzle, a shroud and a bottom spike moving from the top to the bottom. The fuel element consists of eight components in a stainless steel cladding tube, namely the upper and lower end caps, a tighten spring, fuel pellets, upper and lower reflectors, a gas chamber and a counter weight. To ensure the inherent reactivity control, the fuel doppler coefficient and coolant temperature coefficient are designed to be negative. The reactor main design parameters are shown in Table 1, and the neutron parameters are shown in Table 2 (Wang et al., 2016; Wu et al., 2016, 2017). The maximum power of reactor fuel pin is 2.65 kW and that of the assembly is 159.25 kW.
Fig. 2. Core fuel assembly layout of the reactor.
2.2. RELAP5 model A detailed RELEAP5 model of the reactor (Fig. 4) (Wu et al., 2015a,b) was established for the LBE cooled research reactor using RELAP5/MOD 4.0 (Chen et al., 2014). In this model, the core model was simplified with the hottest pin, the hottest assembly and average assembly, adding core power to structure numbered 901/902/903 to simulate core. Four primary heat exchangers (HX1 HX4) were modeled to stand for the PHXs. The heat exchange part of each PHX was divided into 10 sections axially. The secondary side inlet of the heat exchanger was set to a constant temperature and flow value. The outlet side was set to a constant pressure. The secondary loops boundary conditions were represented with time dependent volume (TDV) and timeFig. 3. Configuration of fuel assembly.
Table 1 LBE cooled reactor design parameters.
Fig. 1. LBE-cooled reactor schematic diagram.
Parameters
Value
Unit
Power Core inlet/outlet temperature Primary coolant mass flow Secondary inlet/outlet temperature Secondary loop coolant pressure Secondary loop coolant mass flow Control rods falling time Fuel Number of fuel assemblies Number of primary pumps Primary coolant Secondary coolant
10 300/385 811.67 215/230 4 170 10 UO2 86 2 LBE Pressurized liquid water
MW °C kg/s °C MPa kg/s s – – – – –
dependent junctions (TDJ). The primary pumps (107/117) boundary conditions were represented with TDJ. The nodalization diagram of the RELAP5 model of RVACS is shown in Fig. 5. The air supply and exhaust were represented with two TDVs, components 610 and 650, which were set to the atmospheric pressure. RVACS inlet temperature was set to 50 °C. The RVACS air riser and down-comer tubes were represented by components 620 and 630, respectively.
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Table 2 LBE cooled reactor neutron parameters. Parameters
Value
Unit
Effective delayed neutron fraction Doppler constant Coolant expansion coefficient Axial core expansion coefficient Radial core expansion coefficient
706 0.44 0.53 0.072 0.788
pcm pcm/°C pcm/°C pcm/°C pcm/°C
Scram signal is the power value that increases up to 1.1 times than the normal operation power value (11 MW), and the core outlet temperature that exceeds 395 °C. It is assumed that there is 1.2 s delay for signal transfer of scram command after a trigger for the same is generated. Fuel damage should be found out by studying peak cladding temperature in the core in the process of accident evolution. The standard for fuel cladding damage (referring to European SEVENTH FRAMEWORK PROGRAMME, 2013) is that the cladding temperature must not exceed 800 °C. 3. Transient thermal-hydraulic analysis 3.1. Steady state simulation To study the calculation foundation for accident analysis, the steady state simulation of the LBE reactor was performed at the initial fuel life using RELAP5. The reactor design and simulation values comparisons are shown in Table 3. It shows that the simulation values were in good agreement with design values.
Fig. 5. Nodalization diagram of RELAP5 model of RVACS.
3.2. Accident analysis 3.2.1. Control rod withdrawal accident with scram In this paper, the RIA was studied that some control rods were withdrawn causing the increase of core reactivity. According the reactor design, total reactivity should be less than 0.002 4K/K in
reactivity insertion accident. With regard to the accident, it was assumed that the reactivity was inserted linearly, with a rate of 200.76 pcm in 4 s. The reactor was shut down during the transient accident according to the scram triggered signal.
Fig. 4. Nodalization diagram of RELAP5 model of the reactor.
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415
Table 3 Comparison of simulated values against reactor design parameters. Parameters
Design Value
RELAP5
Difference (%)
Power (MW) Average inlet temperature (°C) Average outlet temperature (°C)
10 300 385
9.70 301 384
3 0.33 0.33
The core power and reactivity changes are shown in Fig. 6. In the beginning, the core power rose fast due to the reactivity insertion. When the core power increased up to 11 MW, the scram signal was triggered. After reactor shut down, the secondary systems stopped following the feed water pumps close. Simultaneously, the decay heat was removed by opening the triggering valve of RVACS system. After 1.2 s delay, the reactor scrammed at 2.2 s. Then the core power decreased rapidly, so did the reactor reactivity. The core power climbed to the maximum value of 12.3 MW at about 2.2 s. The reactivity of the reactor held the maximum value of 108.9 pcm. The coolant inlet, coolant outlet, cladding and fuel temperature changes in first 20 s are presented in Fig. 7. The coolant outlet, cladding and fuel maximum temperature reached a maximum value of about 385 °C, 418 °C and 463 °C, respectively. The cladding peak temperatures stabilized far below 800 °C. The RVACS power surpassed the core decay power at about 2.8 s. After that time the cladding temperature began to decline. The coolant mass flow rate is shown in Fig. 8. With the core power increasing, the temperature difference between coolant outlet and inlet became lager leading to coolant buoyancy effect capacity enhancement. As a result, the coolant mass flow rate rose to remove more heat produced by the core in the first 2.2 s, and then decreased for the reactor shutdown and temperature decline, reducing to 803.90 kg/s at the minimum, with a change of 0.979%. And then it rose and tried to return back to the initial flow rate for the mass regulation of pump represented with TDJ in RELAP5 model.
3.2.2. Control rod withdrawal accident without scram The control rod withdrawal accident without scram has the same reactivity insertion rate as that accident with scram. The reactor did not shut down during this accident by scram system. The core power and reactivity changes are shown in Fig. 9. Fig. 10 presents an enlargement of Fig. 9.The reactor reactivity increased fast during the first 4 s because of the reactivity insertion and then decreased due to the doppler effect, coolant and core expansion negative temperature feedback effect (as shown in Fig. 11). Due to above negative feedback effect, the reactivity
Fig. 6. The core power and reactivity changes in control rod withdrawal accident with scram.
Fig. 7. The core inlet, outlet, fuel, and cladding temperature changes in control rod withdrawal accident with scram.
Fig. 8. The coolant mass flow rate change in control rod withdrawal accident with scram.
Fig. 9. The core power and reactivity change in control rod withdrawal accident without scram.
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Fig. 10. An enlargement of Fig. 9. Fig. 12. The core inlet, outlet, fuel, and cladding temperature change in control rod withdrawal accident without scram.
Fig. 11. Reactivity feedback in control rod withdrawal accident without scram.
climbed to the peak value of 193 pcm rather than 200.76 pcm. After 4 s, the reactor total reactivity decreased rapidly due to the negative feedback effect, with slow increase of the reactor power. The maximum value of the core power was 35.8 MW at 250 s. As time went on, the total reactivity decreased to be negative (less than 0 pcm) at 250 s, causing the core power to go down after then. The coolant inlet, coolant outlet, cladding and fuel temperature change in the short time are presented in Fig. 12. The coolant outlet, cladding and fuel maximum temperature reached a maximum value of about 613 °C, 714 °C and 907 °C, respectively, which were all below the safety design limits. The coolant mass flow rate change is shown in Fig. 13. With the core power increasing, the temperature difference between coolant outlet and inlet becomes larger leading to coolant buoyancy effect capacity enhancement. As a result, the coolant mass flow rate rose to remove more heat produced by the core, increasing to 830.42 kg/s at the maximum, with a flow rate change of 2.31%. And then it tried to decrease to the initial flow rate for the mass regulation of pump represented with TDJ in RELAP5 model. 4. Conclusions A detailed RELAP5 model has been established for accident analysis of the 10 MWth LBE cooled reactor in order to evaluate reactor cladding damage. The model is capable of analyzing a wide
Fig. 13. The coolant mass flow rate change in control rod withdrawal accident without scram.
range of accident scenarios. Two related RIAs, named as control rod withdrawal accident with scram and that without scram, have been simulated. Reactor power and peak cladding temperature have been examined in detail. The reactor power grows exponentially, owing to the constant positive reactivity insertion rate of 50.19 pcm per second, and drops very rapidly after reactor scram. The change in the reactor coolant flow rate regulated by pumps is very small with a maximum change of 2.31%, which is far less than that of the natural circulation reactor. Cladding temperature increases due to the energy deposition during the power rise. The maximum cladding temperature during the control rod withdrawal accident without scram is 714 °C, which is below temperature limit. As a result, there is no damage to fuel cladding in both the control rod withdrawal accident with scram and without scram, which will not lead to the leakage of the radionuclide. Therefore, it is concluded that the RIA will not result in safety violation. By comparison with the two accident analysis results, no fuel cladding damage occurs and inherent safety feature of the 10 MWth LBE cooled reactor is proved owing to its negative feedback.
T. Li et al. / Annals of Nuclear Energy 129 (2019) 412–417
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