scholarly journals Multi-state Risk-Based Maintenance Analysis of Redundant Safety Systems Using the Markov Model and Fault Tree Method

2021 ◽  
Vol 9 ◽  
Author(s):  
F. Mohammadhasani ◽  
A. Pirouzmand
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
System Level ◽  
Aging Effects ◽  
Maintenance Activities ◽  
Maintenance Policies ◽  
Safety Systems ◽  
Tree Method

The risk-based maintenance strategy has received special attention in the safe operation of nuclear power plants. Simultaneous quantification of the positive and negative effects of maintenance activities and components degradation effect makes it possible to accurately evaluate the risk criterion for safety systems of nuclear power plants. However, it is difficult to integrate the effects of maintenance and components degradation into the standard reliability approaches. A straightforward approach for considering components degradation and different maintenance policies is to make use of Markov maintenance models. In this article, the effectiveness of maintenance activities (including changes in the surveillance test intervals and alteration in the different maintenance policies) on the components unavailability with considering aging effects is quantified using Markov maintenance models and then by coupling these models and the fault tree method, the risk measure is upgraded from the component level to the system level. The proposed models are applied to evaluate the unavailability of two safety systems of VVER-1000/V446 nuclear power plants as case studies. The results show that the Markov method due to its multi-state nature is effective in the conservative evaluation of risk measures so that the unavailability computed by the coupling process is higher than the original unavailability (calculated by system fault tree using PSA data of nuclear power plants) for all maintenance policies. In addition, this study illustrates that the developed Markov maintenance models could be applied to the large-scale whole plant level and provides a proper transition from the classical PSA methods to new techniques. This approach integrates the effects of maintenance strategies and components degradation. Also, it provides a practical and a more accurate tool to determine the technical specification of a real nuclear power plant from the risk point of view.

scholarly journals A Method to Avoid Underestimated Risks in Seismic SUPSA and MUPSA for Nuclear Power Plants Caused by Partitioning Events

Energies ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 2150
Author(s):  
Woo Sik Jung
Keyword(s):  
Exact Solutions ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
Ground Acceleration ◽  
Special Operations ◽  
Core Damage ◽  
Full Power ◽  
Conditional Events

Seismic probabilistic safety assessment (PSA) models for nuclear power plants (NPPs) have many non-rare events whose failure probabilities are proportional to the seismic ground acceleration. It has been widely accepted that minimal cut sets (MCSs) that are calculated from the seismic PSA fault tree should be converted into exact solutions, such as binary decision diagrams (BDDs), and that the accurate seismic core damage frequency (CDF) should be calculated from the exact solutions. If the seismic CDF is calculated directly from seismic MCSs, it is drastically overestimated. Seismic single-unit PSA (SUPSA) models have random failures of alternating operation systems that are combined with seismic failures of components and structures. Similarly, seismic multi-unit PSA (MUPSA) models have failures of NPPs that undergo alternating operations between full power and low power and shutdown (LPSD). Their failures for alternating operations are modeled using fraction or partitioning events in seismic SUPSA and MUPSA fault trees. Since partitioning events for one system are mutually exclusive, their combinations should be excluded in exact solutions. However, it is difficult to eliminate the combinations of mutually exclusive events without modifying PSA tools for generating MCSs from a fault tree and converting MCSs into exact solutions. If the combinations of mutually exclusive events are not deleted, seismic CDF is underestimated. To avoid CDF underestimation in seismic SUPSAs and MUPSAs, this paper introduces a process of converting partitioning events into conditional events, and conditional events are then inserted explicitly inside a fault tree. With this conversion, accurate CDF can be calculated without modifying PSA tools. That is, this process does not require any other special operations or tools. It is strongly recommended that the method in this paper be employed for avoiding CDF underestimation in seismic SUPSAs and MUPSAs.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

The application efficiency of intumescent coatings for power cables of safety systems of nuclear power plants under fire conditions

2021 ◽  
Vol 30 (4) ◽  
pp. 36-47
Author(s):  
O. S. Lebedchenko ◽  
S. V. Puzach ◽  
V. I. Zykov
Keyword(s):  
Outer Surface ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Retardant ◽  
Insulation Material ◽  
Power Cables ◽  
Fire Retardants ◽  
Safety Systems ◽  
Fire Conditions

Introduction. The reliable operation of safety systems, that allows for the failure of no more than one safety system component, entails the safe shutdown and cool-down of an NPP reactor in the event of fire. However, the co-authors have not assessed the loss of performance by an insulating material, treated by intumescent compositions and used in the power cables of the above safety systems exposed to the simultaneous effect of various modes of fire and current loads.Goals and objectives. The purpose of the article is the theoretical assessment of the application efficiency of intumescent fire-retardant coatings in power cables used in the safety systems of nuclear power plants having water-cooled and water-moderated reactors under fire conditions. To achieve this goal, the temperature of the outer surface of the insulation and the intumescent fire-retardant coating was analyzed depending on the mode of fire. Theoretical foundations. A non-stationary one-dimensional heat transfer equation is solved to identify the temperature distribution inside the multilayered insulation and the fire-protection layer of a conductive core.Results and their discussion. The co-authors have identified dependences between the temperature of the outer surface of the insulation and the fire retarding composition of the three-core cable VVGng (A)-LS 3x2.5-0.66, on the one hand, and the temperature of the indoor gas environment for three standard modes of fire and one real fire mode. It is found that before the initiation of the process of destruction of the insulation material, the intumescence of the fire-retardant coating occurs only in case of a hydrocarbon fire. Under real fire conditions, the maximal insulation melting time before the initiation of intumescence of the fire-retardant coating at the minimal temperature of intumescence is 4.75 minutes, while the maximal time period from the initiation of destruction of the insulation material to the moment of the insulation melting is 6.0 minutes.Conclusions. An experimental or theoretical substantiation of parameters of intumescent fire retardants, performed using standard modes of fire, has proven the potential loss of operational properties by insulating materials of power cables, used in the safety systems of nuclear power plants, in case of a real fire. Therefore, it is necessary to establish a scientific rationale for the efficient use of fire retardants in the above cables with regard for the conditions of a real fire.

Steady-State Thermal-Hydraulic Analysis of a Particle-Bedded Boiling Water Reactor

2009 ◽  
Author(s):  
Xiaoyu Cai ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Changyou Zhao
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Superheated Steam ◽  
Nuclear Reactor ◽  
Hydraulic Analysis ◽  
Light Water ◽  
Safety Systems ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis

The current Light Water Reactors both BWR and PWR have extensive nuclear reactor safety systems, which provide safe and economical operation of Nuclear Power Plants. During about forty years of operation history the safety systems of Nuclear Power Plants have been upgraded in an evolutionary manner. The cost of safety systems, including large containments, is really high due to a capital cost and a long construction period. These conditions together with a low efficiency of steam cycle for LWR create problems to build new power plants in the USA and in the Europe. An advanced Boiling Water Reactor concept with micro-fuel elements (MFE) and superheated steam promises a radical enhancement of safety and improvement of economy of Nuclear Power Plants. In this paper, a new type of nuclear reactor is presented that consists of a steel-walled tube filled with millions of TRISO-coated fuel particles (Micro-Fuel Elements, MFE) directly cooled by a light-water coolant-moderator. Water is used as coolant that flows from bottom to top through the tube, thereby fluidizing the particle bed, and the moderator water flows in the reverse direction out of the tube. The fuel consists of spheres of about 2.5 mm diameter of UO2 with several coatings of different carbonaceous materials. The external coating of steam cycle the particles is silicon carbide (SiC), manufactured with chemical vapor deposit (CVD) technology. Steady-State Thermal-Hydraulic Analysis aims at providing heat transport capability which can match with the heat generated by the core, so as to provide a set of thermal hydraulic parameters of the primary loop. So the temperature distribution and the pressure losses along the direction of flow are calculated for equilibrium core in this paper. The calculation not only includes the liquid region, but the two phase region and the superheated steam region. The temperature distribution includes both the temperature parameters of micro-fuel elements and the coolant. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.

ACCIDENTS AND INCIDENTS AS A TOOL FOR FINDING THE LOCATIONS AND CAUSES OF FAILURES PROTECTIVE SHELLS FOR NUCLEAR POWER PLANTS IN THE WORLD

2019 ◽  
pp. 14-19 ◽  
Author(s):  
V. V. Potapov ◽  
V. A. Ilin ◽  
A. I. Goshko
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Technical Diagnostics ◽  
Safety Barrier ◽  
The Press ◽  
The World ◽  
Search Tool ◽  
Safety Systems ◽  
Causes Of Failure

In accordance with GOST 20911–89. Technical diagnostics search for the place and causes of failure is the main or one of the main tasks of technical diagnostics of objects. Published information on accidents and incidents is an effective search tool. The base of TOP FAILURES at the nuclear power plants of the world for the 70-year period of operation numbering over 900 cases published in the press is used. The analysis of cases related to the failure of structures, including protective shells, performing the most important function of the final fourth safety barrier of nuclear power plants. Examples of accidents and incidents with protective shells of nuclear power plants with VVER and their safety systems are given.

A verification and validation method and its application to digital safety systems in ABWR nuclear power plants

1998 ◽  
Vol 183 (1-2) ◽  
pp. 117-132 ◽  
Author(s):  
Akira Fukumoto ◽  
Toshifumi Hayashi ◽  
Hiroshi Nishikawa ◽  
Hiroshi Sakamoto ◽  
Teruaki Tomizawa ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Verification And Validation ◽  
Validation Method ◽  
Safety Systems
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