| 1. |
- Kumar, Ranjan, et al.
(författare)
-
A PSA Level-1 method with repairable components : An application to ASTRID Decay Heat Removal systems
- 2015
-
Ingår i: Safety and Reliability. - : CRC Press. - 9781138026810 ; , s. 1611-1617
-
Konferensbidrag (refereegranskat)abstract
- Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I&C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RFT) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP.
|
|
| 2. |
- Kumar, Ranjan, et al.
(författare)
-
A PSA Level-1 method with repairable components : An application to ASTRID Decay Heat Removal systems
- 2014
-
Ingår i: Safety and Reliability: Methodology and Applications. - : CRC Press. ; , s. 1611-1617
-
Bokkapitel (övrigt vetenskapligt/konstnärligt)abstract
- Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I&C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RFT) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP.
|
|
| 3. |
- Curnier, F., et al.
(författare)
-
Symbiosis of static and dynamic probabilistic approaches to support the design process and evaluate the safety of a SFR
- 2015
-
Ingår i: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015. - : American Nuclear Society. - 9781510808119 ; , s. 448-453
-
Konferensbidrag (refereegranskat)abstract
- ASTRID, the Advanced Sodium Technological Reactor for Industrial Demonstration, is a GEN IV technological demonstrator to be commissioned near the end of the 2020 decade. The aim is to demonstrate the progress made in the field of Sodium Fast Reactor technology on an industrial scale, by qualifying innovative options, especially those pertaining to safety and operability. An original combined methodology for probabilistic safety assessment (PSA) is being developed by the CEA and its partners, AREVA NP and EDF at the conceptual design stage of ASTRID. It consists at first, of a static level 1 PSA based on the conventional fault trees (FT)/event trees (ET) approach, taking into account a time period of a week without repair of component malfunctions. Its goal is to provide probabilistic insights in the assessment of design choices and to suppress the weaknesses of the design in terms of safety considerations. A reference configuration of the safety systems is evaluated in order to identify dominant accident sequences. Sensitivity studies are then performed on various design alternatives to define the optimal safety systems configurations that will minimize core damage frequency. It takes into account recent design evolutions for decay heat removal (DHR) systems and support systems, and re-evaluates the preliminary results from ASTRID PSA modeling. The conventional FT/ET approach initially developed for PWRs (Wash 1400) appears to be unsuitable for Sodium Fast Reactors (SFR) PSA because: This approach is binary and static, The probabilistic study for SFR cannot be limited to short periods of time - when repair is not possible - because several months are necessary for the thermal leakage to be equivalent to decay heat, SFR technology cannot rely simply on DHR complementary systems, The modeling by FT/ET is not designed for long periods of time, Repair, on along and middle term basis, of failed components is not considered. Therefore, dynamic PSA approaches have been investigated to extend the conventional PSA to longer periods of time by taking into account the specific characteristics of a sodium reactor such as its great thermal inertia - which allows the operator to make interventions - and the fact that sodium circuits present risks of irreversible and temperature-sensitive failures. What these approaches have in common is the possibility of taking into account the repair of failed components. Simplified thermal-hydraulic calculations were performed to characterize the reactor at any given moment in the accident scenario. The benefits of dynamic approaches on short periods of time will be quantitatively evaluated in 2015.
|
|
| 4. |
- Kumar, Ranjan, et al.
(författare)
-
Dynamic Hybrid Reliability Studies of a Decay Heat Removal System
- 2015
-
Ingår i: International Journal of Reliability, Quality and Safety Engineering (IJRQSE). - : World Scientific. - 0218-5393. ; 22:4
-
Tidskriftsartikel (refereegranskat)abstract
- Some critical safety systems exhibit the characteristics of hybrid stochastic class whose performance depends on the dynamic interactions of deterministic variables of physical phenomena and probabilistic variables of system failures. However, conventional probabilistic safety assessment (PSA) method involves static event and linked fault tree analysis and does not capture the dynamic interactions of such hybrid stochastic systems. Additionally, the existing dynamic PSA methods do not consider any repair possibility of some failed components during safety assessment. To address these issues, this paper presents a dynamic hybrid reliability assessment scheme for performance studies of repairable nuclear safety systems during a mission time. This scheme combines the features of reliability block diagram (RBD) for system compositions and partial differential equations for system physics using a customized stochastic hybrid automata tool implemented on Python platform. A case study of decay heat removal (DHR) systems has been performed using the introduced scheme. The impacts of failure rates and repair rates on sodium temperature evolution over a mission time have been analyzed. The results provide useful safety insights in mission safety tests of DHR systems. In sum, this work advances the dynamic safety assessment approach for complex system designs including nuclear power plants.
|
|
