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Design failure mode and effect analysis for Korean fusion DEMO plant
Fusion Engineering and Design 87 (2012) 412–417
Contents lists available at SciVerse ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Design failure mode and effect analysis for Korean fusion DEMO plant Myoung-suk Kang a , Jeongtae Cho a , Rizwan Ahmed a , Gyunyoung Heo a,∗ , Hyuck Jong Kim b , Young-seok Lee b a b
Kyung Hee University, Yongin-si, Gyeonggi-do 446-701, Republic of Korea National Fusion Research Institute, Daejeon-si 305-333, Republic of Korea
a r t i c l e
i n f o
Article history: Available online 14 December 2011 Keywords: Korean fusion DEMO plant Failure mode and effect analysis Safety analysis Simulation, Safety features
a b s t r a c t The construction of Korean Fusion DEMO Plant (KFDP) for a demonstration of the plasma analysis and engineering feasibilities is planned in 2030s based on the Korean fusion technology roadmap. The radiation safety should be assured for nuclear facilities, so that, the KFDP is required to research for the regulatory requirements and industrial codes and standards. The final design guidance of the engineered safety features should be served in future. As the first step for this research, the failure modes and effects analysis in a design stage was performed. This leads to find the list of potential hazard elements and to obtain the list of initiating events for the future probabilistic risk assessment. The hazard elements expected to seriously threaten the integrity of the KFDP were investigated and determined to quantify the effect of the initiating events in the effect analysis: (1) a total loss of active cooling water to occur during the burn with decay heat calculation and (2) coolant ingress from cooling circuits into the vacuum vessel, cryostat and containment building. For those initiating events, the quantitative simulations using transient mass and energy calculation and computational fluid dynamics were performed. © 2011 Elsevier B.V. All rights reserved.
1. Introduction In accordance with the Korean fusion technology roadmap, the construction of demonstration fusion power plants is scheduled for the investigation of technical and commercial feasibilities around 2030. The National Fusion Research Institute (NFRI) in South Korea has initiated the engineering R&D programs to make fusion technologies practical forward in 2010. Currently, NFRI has been developing the function tree that is to hierarchically present the bare-bone systems related to electricity generation using Tokamak and power conversion systems to embrace a wide range of issues and uncertainties in selecting design parameters. The operating conditions of the designed system are similar to the existing commercial fission-based power plant, specifically, thermal power, electric power and plant efficiency. Also, it has the comparable power conversion system with a pressurized water reactor’s system. The terminals in the function tree correspond to the design parameters and their failure modes have influence on the performance as well as on the safety of the entire power plant [1]. This paper provides the preliminary results on the Failure Mode and Effects Analysis (FMEA) at the conceptual design stage; we call it Design-FMEA (DFMEA), which is to explore the relevant Engi-
∗ Corresponding author. Tel.: +82 31 204 3835. E-mail address: [email protected] (G. Heo). 0920-3796/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2011.11.007
neering Safety Features (ESFs) of Korean Fusion DEMO Plant (KFDP), and that would ultimately lead to the establishment of the design and regulatory requirements of ESFs for commercial fusion-based power plants. The results of failure mode analysis are presented as a list of potential hazard elements and their failure frequency, which will contribute to the future probabilistic risk assessment. Some hazard elements expected to threaten seriously were represented to quantify the effects of the initiating events. The effects of failures were evaluated in a qualitative manner through engineering judgment. For the initiating events seriously considered from the radiological safety viewpoint, the quantitative simulations were also performed to decide the requirements of the associated ESFs.
2. Methodology The main purpose of this study is to perform DFMEA for the ESFs of KFDP, which ultimately aims at establishing the design and regulatory requirements of the ESFs. In previous studies [1], we conducted researches in deterministic and probabilistic safety analyzes to take the synergetic benefits of both methodologies in achieving safety goals. In addition, we developed a function tree which hierarchically presented the bare-bone systems related to electricity generation using Tokamak and power conversion systems. Although the research on the conceptual design and technical safety issues for KFDP has partially been done, the investigation on
M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
413
and water-cooled solid type blankets. This DEMO power plant is expected to have same size as ITER but with higher plasma density. In addition to this plan, we assumed following factors for DFMEA:
1. 30% thermal efficiency, 80% capacity factor, and 360 Effective Full Power Day (EFPD)/1 year 2. Homogeneous plasma in a Vacuum Vessel (VV) 3. No neutron irradiation to the magnet system 4. Same material and inventory with ITER’s Tokamak system and other subsidiary systems
Fig. 1. DFMEA procedure for safety analysis of KFDP.
failure modes and effects is required to adjust as plans for KFDP have modified. The implementation plans for the R&D Program have been established as they were pulled from the KFDP of which the total capacity will be around 600 MWe with a supercritical Rankine cycle
Adopting above assumptions, at first we modified the function tree in the form of the bare-bone systems from the preceding study [1]. This design process is suitable when the level of the plant design is not so detailed and it justifies FMEA directly at component or system level. Fig. 1 simply explains our procedure for the DFMEA. The design parameters at the bottom of the function tree and their failure modes affect the performance as well as the safety on the entire power plant. Failures of the terminal design parameter are basic events in the failure domain [1]. From the viewpoint of the safety analysis, investigating the propagation of a basic event is required to consider whether the basic event would bring the entire system smoothly to the shutdown state or it would cause a seriously abnormal situation. The later cases we focus on are initiating events at the bottom of incidents or accidents. To select the list of initiating events for an individual design parameter, the failure mode analysis should be conducted, and their effects should also be evaluated in a qualitative manner by engineering judgment.
Fig. 2. Specific activity after reactor shut down after 360 EFPD (Upside: outboard/Downside: inboard).
414
M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
Table 1 The result of FMEA for KFDP 600 MWe. Design Parameters
Failure modes
Effects
DP 1.1.1
Heating system failure (ECH, ICRF, NBI, LHCD) Structural failure Multiplication failure
Loss of plasma control
Vertical displacement event
Loss of Vacuum Excess tritium H2 combustion (Be multiplier + steam)
H2 detonation Radiative material leak from 1st, 2nd confinement
T breading failure Vacuum pumping system failure
Loss of primary vacuum pump boundary
Structural failure
Loss of control of effects of magnetic energy (magnet shift impacts VV, arc to VV, and missile impacts VV) Failure in ground insulation of coils, cryogenic services and structures Coolant ingress Excessive heat in leakage/He leak Pump shaft/Valve leak
Vertical displacement event
Loss of control of hydrogen inventories & chemical reaction Air ingress Hot gas admission Pump trip/Pump leak/break Pipe & Valve leak, break/Lock/Open failure Loss of control Storage leak/break
Radioactive material leak from 1st, 2nd confinement H2 detonation
DP 1.1.2
DP 1.1.3
Coil control failure Partial coil failure Loss of power control Power supply system failure DP 1.1.4
Structural failure Loss of electrical power Fail to run
Radioactive material leak from 1st, 2nd confinement He leak from 2nd confinement
DP 1.1.5
Start-up heat removal system failure
Loss of heat removal
Radioactive material leak
DP 1.2.4
Structural failure System failure
Loss of heat removal Loss of flow Loss of heat sink In-cryostat coolant ingress Ex-vessel coolant ingress
H2 detonation Vertical displacement event Radioactive material leak from 1st, 2nd confinement
DP 1.3.1
Structural failure System failure
DP 1.3.1
DP 1.3.3
Structural failure
Simultaneous coolant ingress in VV and vault Simultaneous coolant ingress into plasma chamber and cryostat Plasma overpower Detritiation system failure Isotope separation system failure Loss of heat sink Delivery cut-off system failure
Power termination system failure
Fast impurity gas injection system failure
Structural failure
Pump trip
Exhaust system failure
Pump leak/break
3. Design failure modes & effects analysis The results of FMEA considered in this study are limited to internal initiators that are assumed to occur during the normal operation of KFDP. The purpose of the function tree was to get a feasible design for satisfying the only performance-related requirement, that is, electric power generation. Therefore, the weakness from the viewpoint of system’s reliability should be analyzed for improving system’s safety by means of accident prevention and/or mitigation. A preliminary analysis of the FMEA was based on the function tree designed on the basis of open innovation and optimal design process considering all of the potential options from a conceptual design stage [1]. Thus the candidate lists of initiating events have been modified according to the real condition of KFDP’s function tree. Assuming the failures of each terminal design parameter in this function tree, the fault tree is able to be obtained by complementary conversion of the function tree. The lowest level of fault tree consists of basic events which cause system failures. Through assessing the effects of basic events, the events able to develop as accidents or incidents were elicited as initiating events. The list of initiating events and their effects is provided in Table 1. We screened, grouped
H2 detonation Vertical displacement event Radioactive material leak from 1st, 2nd confinements Radioactive material leak from 1st, 2nd confinement Radioactive material leak from 1st, 2nd confinement
and finalized this list according to their severity and physical phenomena. The right side of the effects in this table presents the most severe cases caused by left side effects in the same line. According to this result, the hazard elements expected to threaten most seriously were represented to quantify the effect of the initiating events: (1) a total loss of active cooling water to occur during the burn, and (2) coolant ingress from blanket cooling pipe into the vessel. Analysis of those events performed to assess the requirements of the associated ESFs. The detailed description of this analysis and required design data is reported in following sections.
4. Case #1: a total loss of active cooling water to occur during the burn This effect analysis is assumed to be a total loss of cooling flow from all loops. The heat rejection is only by passive conduction through the layers, towards the outer regions where eventually a heat sink is provided by convective circulation of the atmosphere in a building.
M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
415
Table 2 Analysis result for cooling loop structural failures by MARS code. Containment Size
1/2
2
4
8
Max. released rate (kg/s) Max. Pressure (bar) Max. temp (K)
9.65 4.05 402.83
147.74 4.03 403.35
555.87 4.06 409.91
1898.2 4.14 418.76
Cryostat Max. released rate (kg/s) Max. pressure (bar) Max. temp (K)
9.65 9.63 444.23
147.76 9.65 450.55
555.94 9.8 460.04
1898 9.94 469.91
Vacuum vessel Max. released rate (kg/s) Max. pressure (bar) Max. temp (K)
9.65 40.5 528.41
147.74 41.55 548.69
555.99 41.64 553.89
1898 41.39 557.16
Fig. 3. Model configurations for temperature transient simulation.
4.1. Decay heat calculation The amount of decay heat generation in a fusion reactor after a 360 EFPD was computed using the following methodology. The problem simulated was to estimate the amount of heat generated in case an accident occurs at the end of power generation cycle. The production and destruction of a radionuclide during reactor operation in structural materials of fusion reactor such as blanket, VV, and divertor were considered to contribute to the decay heat. A nuclide’s production and decay scheme under various nuclear reactions is governed by the following equation.
dXi = lij j Xj + k Xk − (i + i )Xi dt N
N
j=1
k=1
where Xi = atom density for nuclide i; lij = decay fraction of nuclide j to nuclide i; i = decay constant of nuclide i; i = spectrum-averaged neutron capture cross-section for nuclide i; = one-group neutron flux. The equation considers the influences of neutron flux, compositions of materials, nuclides capture cross-sections and their half-lives on the activity generation and therefore, provides a reasonable estimate for the amount of radioactive materials. However, in many cases it was very difficult to provide all the required parameters of Eq. (1) perfectly, specifically the fluxes and cross sections due to geometrical complexity and material heterogeneity. For the reason stated above, the problem of nuclide buildup and decay was solved with a two-step approach involving computation of neutron distribution in the first step followed by the solution of nuclide buildup and decay equation [2]. A combination of computer codes from reactor physics and nuclide inventory are usually employed such calculations. In the first step, we computed the neutron flux profile and one-group cross sections for important reactions using MCNP simulation. MCNP4c3 code [3] was used to give the flux degradation and spectrum information in structural materials of the fusion reactor. In this step, a simplified model for the geometry was considered to obtain reasonable estimates for the quantities of interest. The extension of the structure into the plane of paper was assumed to be 65.9 m, which was essential to equate the plasma volumes in the model and ITER. In MCNP simulation
(1)
Fig. 4. Temperature transient histories following onset of hypothetical total loss of cooling flow.
Fig. 5. Schematic diagram for coolant ingress simulation.
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M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
materials. The computation of one-group reaction cross sections in this way to reflect the true reactor environment was important for the forthcoming calculations, since these cross sections were to be utilized in the second step [5]. Eq. (2) was used to obtain the cross sections from MCNP-calculated quantities. ¯ =
(E)(E)dEdV (E)dEdV
(2)
where ¯ = one-group collapsed cross-section i; (E) = energy dependent capture cross-section; (E) = energy dependent flux. In the second step, nuclide buildup and destruction under the influence of neutron flux was computed using ORIGEN2 code [6], where important one-group reaction cross sections computed from Eq. (2) were supplied to its cross section library. The results of decay heat computed form MCNP/ORIGEN2 system are shown in Fig. 2 for out-board and in-board structural components, respectively. These figures represent the total amount of specific activity (Bq/kg) attributed to various neutron induced reactions such as (n,), (n,n), (n,p), etc. in the layers of structural materials as a function of time. The decay was only observed to drop rapidly in Be-coating in first five hours it was observed to remain on the order of tens of kilowatt after 15 h. The decay heat was about tens of kilowatt with highest values as 265.6, 110.4 kW in out-board divertor and blanket at the time of accident (at 360 EFPD), respectively. Similarly, the highest decay heat of 79.6 kW was observed in Blanket region among the in-board structural materials. 4.2. Temperature transient calculations To evaluate the quality of structures in 1st confinements, temperature transients were computed in Fig. 3. These calculations were performed using ANSYS workbench 11.0 code [7] with a 3-D solid geometry model similar to one used for the decay heat calculation. In accordance with the no active cooling in the model, the water coolant was assumed to be absent. This case study is based on a conservative set of assumptions and may not correspond to a real situation, but it is an attempt to provide a clearly bounding condition for the most extreme cases. The result from this calculation is shown in Fig. 4. From this analysis, the temperature limitation for 1st confinement’s structures was roughly grasped. If the melting point of the structures is below the maximum temperatures, some ESFs for cooling would be required at early stage of temperature transient, for instance, emergency tokamak cooling systems. 5. Case #2: coolant ingress from cooling pipe into the vessel, cryostat and containment building
Fig. 6. Pressure transient inside VV, cryostat, containment building with isolation valves.
step, we computed the reference flux profile and one-group cross sections for important reactions such as (n,) and (n,˛) in structural components. The reference flux profile was then converted to absolute flux by using the typical flux value of 0.57 MW/m2 [4], and the flux ratio in the reference flux profile. The absolute flux was highest as 2.52 × 1013 n/cm2 s on Be-Coating and was lowest as 1.69 × 107 n/cm2 s on Toroidal Field (TF) coil marking an overall degradation on the order of 106 in neutron flux in structural
In order to assess the severity of initiating events related to coolant leakage, we attempted to analyze loss of coolant accidents (LOCAs) occurred inside and outside a VV. In these cases, the large pressure difference between the structure and the coolant pipes will produce large amount of steam and it will build up great pressure on the inner wall of the structures. Eventually, this pressure rise can threaten the integrity of the structures. For that reason, we analyzed 1-dimenonsional temperature/pressure transition using MARS code [8]. The simple scheme for this calculation is showing Fig. 5, and the result of calculation is shown in Table 2. This analysis indicated that the highest pressure level in VV was about 41.39 bar in case of 8 break in blanket cooling pipe. Unless the VV is able to endure the pressure, it requires safety features such as rupture disks, bleed line valves or/and isolation valves. In the same manner, other structures also require safety features. As a simple example, we adopted the isolation valve actuation system
M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
in the 8 break case for all structures. Fig. 6 shows the simulation results assuming isolating capability. 6. Conclusion This study is expected to provide backup materials to prioritize the associated future research on the Korean fusion technology roadmap and other fusion power projects. This research can be regarded as an attempt for the development of regulatory systems and the calculation related to radiation intensity. The results from the failure modes analysis will be provided for the entire direction of KFDP, which can reduce the design weaknesses and unnecessary processes. In the same manner, effects analysis results are able to play the roles as guidelines for technical developments in the fusion area. Acknowledgments This work was supported by R&D Program through the National Fusion Research Institute of Korea (NFRI) funded by the Government funds.
417
References [1] M. Kang, G. Heo, Y. Lee, H. Kim, Study on conceptual design and technical safety issues for korean demonstration fusion reactors, Fusion Science and Technology 60 (1) (2011) 1–8. [2] IAEA, Radiological Characterization of Shutdown Nuclear Reactors for Decommissioning, Technical Report Series No. 389, International Atomic Energy Agency, Vienna, 1998. [3] J.F. Briesmeister, MCNPTM -A General Monte Carlo N-particle Transport Code, Version 4C, LA-13709-M, Los Alamos National Laboratory, USA, 2000. [4] A. Fouossat, ITER Overview, International School on Fusion Technologies, Karlsruhe, 2010, September 06–17. [5] M.A. Abdou, N.B. Morley, A.Y. Ying, S. Smolentsev, P. Calderoni, Overview of Fusion Blanket R&D in the US Over the Last Decade Nuclear Engineering and Technology, Vol. 3, Korean Nuclear Society, 2005, pp. 401–422. [6] D.K. Cho, H.J. Choi, R. Ahmed, G.Y. Heo, Radiological Characteristics of Decommissioning Waste from a CANDU Reactor, Nuclear Engineering and Technology, Korean Nuclear Society, accepted for publication. [7] ANSYS Inc., Documentation for ANSYS Workbench 11.0, 2006. [8] J.J. Jeong, K.S. Ha, B.D. Chung, W.J. Lee, Development of a multi-dimensional thermal–hydraulic system code, MARS 1.3.1, Ann. Nucl. Energy 26 (1999) 1611–1642.
Contents lists available at SciVerse ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Design failure mode and effect analysis for Korean fusion DEMO plant Myoung-suk Kang a , Jeongtae Cho a , Rizwan Ahmed a , Gyunyoung Heo a,∗ , Hyuck Jong Kim b , Young-seok Lee b a b
Kyung Hee University, Yongin-si, Gyeonggi-do 446-701, Republic of Korea National Fusion Research Institute, Daejeon-si 305-333, Republic of Korea
a r t i c l e
i n f o
Article history: Available online 14 December 2011 Keywords: Korean fusion DEMO plant Failure mode and effect analysis Safety analysis Simulation, Safety features
a b s t r a c t The construction of Korean Fusion DEMO Plant (KFDP) for a demonstration of the plasma analysis and engineering feasibilities is planned in 2030s based on the Korean fusion technology roadmap. The radiation safety should be assured for nuclear facilities, so that, the KFDP is required to research for the regulatory requirements and industrial codes and standards. The final design guidance of the engineered safety features should be served in future. As the first step for this research, the failure modes and effects analysis in a design stage was performed. This leads to find the list of potential hazard elements and to obtain the list of initiating events for the future probabilistic risk assessment. The hazard elements expected to seriously threaten the integrity of the KFDP were investigated and determined to quantify the effect of the initiating events in the effect analysis: (1) a total loss of active cooling water to occur during the burn with decay heat calculation and (2) coolant ingress from cooling circuits into the vacuum vessel, cryostat and containment building. For those initiating events, the quantitative simulations using transient mass and energy calculation and computational fluid dynamics were performed. © 2011 Elsevier B.V. All rights reserved.
1. Introduction In accordance with the Korean fusion technology roadmap, the construction of demonstration fusion power plants is scheduled for the investigation of technical and commercial feasibilities around 2030. The National Fusion Research Institute (NFRI) in South Korea has initiated the engineering R&D programs to make fusion technologies practical forward in 2010. Currently, NFRI has been developing the function tree that is to hierarchically present the bare-bone systems related to electricity generation using Tokamak and power conversion systems to embrace a wide range of issues and uncertainties in selecting design parameters. The operating conditions of the designed system are similar to the existing commercial fission-based power plant, specifically, thermal power, electric power and plant efficiency. Also, it has the comparable power conversion system with a pressurized water reactor’s system. The terminals in the function tree correspond to the design parameters and their failure modes have influence on the performance as well as on the safety of the entire power plant [1]. This paper provides the preliminary results on the Failure Mode and Effects Analysis (FMEA) at the conceptual design stage; we call it Design-FMEA (DFMEA), which is to explore the relevant Engi-
∗ Corresponding author. Tel.: +82 31 204 3835. E-mail address: [email protected] (G. Heo). 0920-3796/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2011.11.007
neering Safety Features (ESFs) of Korean Fusion DEMO Plant (KFDP), and that would ultimately lead to the establishment of the design and regulatory requirements of ESFs for commercial fusion-based power plants. The results of failure mode analysis are presented as a list of potential hazard elements and their failure frequency, which will contribute to the future probabilistic risk assessment. Some hazard elements expected to threaten seriously were represented to quantify the effects of the initiating events. The effects of failures were evaluated in a qualitative manner through engineering judgment. For the initiating events seriously considered from the radiological safety viewpoint, the quantitative simulations were also performed to decide the requirements of the associated ESFs.
2. Methodology The main purpose of this study is to perform DFMEA for the ESFs of KFDP, which ultimately aims at establishing the design and regulatory requirements of the ESFs. In previous studies [1], we conducted researches in deterministic and probabilistic safety analyzes to take the synergetic benefits of both methodologies in achieving safety goals. In addition, we developed a function tree which hierarchically presented the bare-bone systems related to electricity generation using Tokamak and power conversion systems. Although the research on the conceptual design and technical safety issues for KFDP has partially been done, the investigation on
M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
413
and water-cooled solid type blankets. This DEMO power plant is expected to have same size as ITER but with higher plasma density. In addition to this plan, we assumed following factors for DFMEA:
1. 30% thermal efficiency, 80% capacity factor, and 360 Effective Full Power Day (EFPD)/1 year 2. Homogeneous plasma in a Vacuum Vessel (VV) 3. No neutron irradiation to the magnet system 4. Same material and inventory with ITER’s Tokamak system and other subsidiary systems
Fig. 1. DFMEA procedure for safety analysis of KFDP.
failure modes and effects is required to adjust as plans for KFDP have modified. The implementation plans for the R&D Program have been established as they were pulled from the KFDP of which the total capacity will be around 600 MWe with a supercritical Rankine cycle
Adopting above assumptions, at first we modified the function tree in the form of the bare-bone systems from the preceding study [1]. This design process is suitable when the level of the plant design is not so detailed and it justifies FMEA directly at component or system level. Fig. 1 simply explains our procedure for the DFMEA. The design parameters at the bottom of the function tree and their failure modes affect the performance as well as the safety on the entire power plant. Failures of the terminal design parameter are basic events in the failure domain [1]. From the viewpoint of the safety analysis, investigating the propagation of a basic event is required to consider whether the basic event would bring the entire system smoothly to the shutdown state or it would cause a seriously abnormal situation. The later cases we focus on are initiating events at the bottom of incidents or accidents. To select the list of initiating events for an individual design parameter, the failure mode analysis should be conducted, and their effects should also be evaluated in a qualitative manner by engineering judgment.
Fig. 2. Specific activity after reactor shut down after 360 EFPD (Upside: outboard/Downside: inboard).
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M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
Table 1 The result of FMEA for KFDP 600 MWe. Design Parameters
Failure modes
Effects
DP 1.1.1
Heating system failure (ECH, ICRF, NBI, LHCD) Structural failure Multiplication failure
Loss of plasma control
Vertical displacement event
Loss of Vacuum Excess tritium H2 combustion (Be multiplier + steam)
H2 detonation Radiative material leak from 1st, 2nd confinement
T breading failure Vacuum pumping system failure
Loss of primary vacuum pump boundary
Structural failure
Loss of control of effects of magnetic energy (magnet shift impacts VV, arc to VV, and missile impacts VV) Failure in ground insulation of coils, cryogenic services and structures Coolant ingress Excessive heat in leakage/He leak Pump shaft/Valve leak
Vertical displacement event
Loss of control of hydrogen inventories & chemical reaction Air ingress Hot gas admission Pump trip/Pump leak/break Pipe & Valve leak, break/Lock/Open failure Loss of control Storage leak/break
Radioactive material leak from 1st, 2nd confinement H2 detonation
DP 1.1.2
DP 1.1.3
Coil control failure Partial coil failure Loss of power control Power supply system failure DP 1.1.4
Structural failure Loss of electrical power Fail to run
Radioactive material leak from 1st, 2nd confinement He leak from 2nd confinement
DP 1.1.5
Start-up heat removal system failure
Loss of heat removal
Radioactive material leak
DP 1.2.4
Structural failure System failure
Loss of heat removal Loss of flow Loss of heat sink In-cryostat coolant ingress Ex-vessel coolant ingress
H2 detonation Vertical displacement event Radioactive material leak from 1st, 2nd confinement
DP 1.3.1
Structural failure System failure
DP 1.3.1
DP 1.3.3
Structural failure
Simultaneous coolant ingress in VV and vault Simultaneous coolant ingress into plasma chamber and cryostat Plasma overpower Detritiation system failure Isotope separation system failure Loss of heat sink Delivery cut-off system failure
Power termination system failure
Fast impurity gas injection system failure
Structural failure
Pump trip
Exhaust system failure
Pump leak/break
3. Design failure modes & effects analysis The results of FMEA considered in this study are limited to internal initiators that are assumed to occur during the normal operation of KFDP. The purpose of the function tree was to get a feasible design for satisfying the only performance-related requirement, that is, electric power generation. Therefore, the weakness from the viewpoint of system’s reliability should be analyzed for improving system’s safety by means of accident prevention and/or mitigation. A preliminary analysis of the FMEA was based on the function tree designed on the basis of open innovation and optimal design process considering all of the potential options from a conceptual design stage [1]. Thus the candidate lists of initiating events have been modified according to the real condition of KFDP’s function tree. Assuming the failures of each terminal design parameter in this function tree, the fault tree is able to be obtained by complementary conversion of the function tree. The lowest level of fault tree consists of basic events which cause system failures. Through assessing the effects of basic events, the events able to develop as accidents or incidents were elicited as initiating events. The list of initiating events and their effects is provided in Table 1. We screened, grouped
H2 detonation Vertical displacement event Radioactive material leak from 1st, 2nd confinements Radioactive material leak from 1st, 2nd confinement Radioactive material leak from 1st, 2nd confinement
and finalized this list according to their severity and physical phenomena. The right side of the effects in this table presents the most severe cases caused by left side effects in the same line. According to this result, the hazard elements expected to threaten most seriously were represented to quantify the effect of the initiating events: (1) a total loss of active cooling water to occur during the burn, and (2) coolant ingress from blanket cooling pipe into the vessel. Analysis of those events performed to assess the requirements of the associated ESFs. The detailed description of this analysis and required design data is reported in following sections.
4. Case #1: a total loss of active cooling water to occur during the burn This effect analysis is assumed to be a total loss of cooling flow from all loops. The heat rejection is only by passive conduction through the layers, towards the outer regions where eventually a heat sink is provided by convective circulation of the atmosphere in a building.
M.-s. Kang et al. / Fusion Engineering and Design 87 (2012) 412–417
415
Table 2 Analysis result for cooling loop structural failures by MARS code. Containment Size
1/2
2
4
8
Max. released rate (kg/s) Max. Pressure (bar) Max. temp (K)
9.65 4.05 402.83
147.74 4.03 403.35
555.87 4.06 409.91
1898.2 4.14 418.76
Cryostat Max. released rate (kg/s) Max. pressure (bar) Max. temp (K)
9.65 9.63 444.23
147.76 9.65 450.55
555.94 9.8 460.04
1898 9.94 469.91
Vacuum vessel Max. released rate (kg/s) Max. pressure (bar) Max. temp (K)
9.65 40.5 528.41
147.74 41.55 548.69
555.99 41.64 553.89
1898 41.39 557.16
Fig. 3. Model configurations for temperature transient simulation.
4.1. Decay heat calculation The amount of decay heat generation in a fusion reactor after a 360 EFPD was computed using the following methodology. The problem simulated was to estimate the amount of heat generated in case an accident occurs at the end of power generation cycle. The production and destruction of a radionuclide during reactor operation in structural materials of fusion reactor such as blanket, VV, and divertor were considered to contribute to the decay heat. A nuclide’s production and decay scheme under various nuclear reactions is governed by the following equation.
dXi = lij j Xj + k Xk − (i + i )Xi dt N
N
j=1
k=1
where Xi = atom density for nuclide i; lij = decay fraction of nuclide j to nuclide i; i = decay constant of nuclide i; i = spectrum-averaged neutron capture cross-section for nuclide i; = one-group neutron flux. The equation considers the influences of neutron flux, compositions of materials, nuclides capture cross-sections and their half-lives on the activity generation and therefore, provides a reasonable estimate for the amount of radioactive materials. However, in many cases it was very difficult to provide all the required parameters of Eq. (1) perfectly, specifically the fluxes and cross sections due to geometrical complexity and material heterogeneity. For the reason stated above, the problem of nuclide buildup and decay was solved with a two-step approach involving computation of neutron distribution in the first step followed by the solution of nuclide buildup and decay equation [2]. A combination of computer codes from reactor physics and nuclide inventory are usually employed such calculations. In the first step, we computed the neutron flux profile and one-group cross sections for important reactions using MCNP simulation. MCNP4c3 code [3] was used to give the flux degradation and spectrum information in structural materials of the fusion reactor. In this step, a simplified model for the geometry was considered to obtain reasonable estimates for the quantities of interest. The extension of the structure into the plane of paper was assumed to be 65.9 m, which was essential to equate the plasma volumes in the model and ITER. In MCNP simulation
(1)
Fig. 4. Temperature transient histories following onset of hypothetical total loss of cooling flow.
Fig. 5. Schematic diagram for coolant ingress simulation.
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materials. The computation of one-group reaction cross sections in this way to reflect the true reactor environment was important for the forthcoming calculations, since these cross sections were to be utilized in the second step [5]. Eq. (2) was used to obtain the cross sections from MCNP-calculated quantities. ¯ =
(E)(E)dEdV (E)dEdV
(2)
where ¯ = one-group collapsed cross-section i; (E) = energy dependent capture cross-section; (E) = energy dependent flux. In the second step, nuclide buildup and destruction under the influence of neutron flux was computed using ORIGEN2 code [6], where important one-group reaction cross sections computed from Eq. (2) were supplied to its cross section library. The results of decay heat computed form MCNP/ORIGEN2 system are shown in Fig. 2 for out-board and in-board structural components, respectively. These figures represent the total amount of specific activity (Bq/kg) attributed to various neutron induced reactions such as (n,), (n,n), (n,p), etc. in the layers of structural materials as a function of time. The decay was only observed to drop rapidly in Be-coating in first five hours it was observed to remain on the order of tens of kilowatt after 15 h. The decay heat was about tens of kilowatt with highest values as 265.6, 110.4 kW in out-board divertor and blanket at the time of accident (at 360 EFPD), respectively. Similarly, the highest decay heat of 79.6 kW was observed in Blanket region among the in-board structural materials. 4.2. Temperature transient calculations To evaluate the quality of structures in 1st confinements, temperature transients were computed in Fig. 3. These calculations were performed using ANSYS workbench 11.0 code [7] with a 3-D solid geometry model similar to one used for the decay heat calculation. In accordance with the no active cooling in the model, the water coolant was assumed to be absent. This case study is based on a conservative set of assumptions and may not correspond to a real situation, but it is an attempt to provide a clearly bounding condition for the most extreme cases. The result from this calculation is shown in Fig. 4. From this analysis, the temperature limitation for 1st confinement’s structures was roughly grasped. If the melting point of the structures is below the maximum temperatures, some ESFs for cooling would be required at early stage of temperature transient, for instance, emergency tokamak cooling systems. 5. Case #2: coolant ingress from cooling pipe into the vessel, cryostat and containment building
Fig. 6. Pressure transient inside VV, cryostat, containment building with isolation valves.
step, we computed the reference flux profile and one-group cross sections for important reactions such as (n,) and (n,˛) in structural components. The reference flux profile was then converted to absolute flux by using the typical flux value of 0.57 MW/m2 [4], and the flux ratio in the reference flux profile. The absolute flux was highest as 2.52 × 1013 n/cm2 s on Be-Coating and was lowest as 1.69 × 107 n/cm2 s on Toroidal Field (TF) coil marking an overall degradation on the order of 106 in neutron flux in structural
In order to assess the severity of initiating events related to coolant leakage, we attempted to analyze loss of coolant accidents (LOCAs) occurred inside and outside a VV. In these cases, the large pressure difference between the structure and the coolant pipes will produce large amount of steam and it will build up great pressure on the inner wall of the structures. Eventually, this pressure rise can threaten the integrity of the structures. For that reason, we analyzed 1-dimenonsional temperature/pressure transition using MARS code [8]. The simple scheme for this calculation is showing Fig. 5, and the result of calculation is shown in Table 2. This analysis indicated that the highest pressure level in VV was about 41.39 bar in case of 8 break in blanket cooling pipe. Unless the VV is able to endure the pressure, it requires safety features such as rupture disks, bleed line valves or/and isolation valves. In the same manner, other structures also require safety features. As a simple example, we adopted the isolation valve actuation system
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in the 8 break case for all structures. Fig. 6 shows the simulation results assuming isolating capability. 6. Conclusion This study is expected to provide backup materials to prioritize the associated future research on the Korean fusion technology roadmap and other fusion power projects. This research can be regarded as an attempt for the development of regulatory systems and the calculation related to radiation intensity. The results from the failure modes analysis will be provided for the entire direction of KFDP, which can reduce the design weaknesses and unnecessary processes. In the same manner, effects analysis results are able to play the roles as guidelines for technical developments in the fusion area. Acknowledgments This work was supported by R&D Program through the National Fusion Research Institute of Korea (NFRI) funded by the Government funds.
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