Reactor Cavity Analysis Under Steam Explosive Conditions by TNT Model

2016 ◽  
Author(s):  
Seung-Huyn Kim ◽  
Yoon-Suk Chang ◽  
Yong-Jin Cho
Keyword(s):  
Shock Wave ◽  
Steam Explosion ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Vessel ◽  
Dynamic Effects ◽  
Element Mesh ◽  
Cooling Strategy ◽  
Hydraulic Conditions ◽  
Physical Phenomena

Steam explosion may occur in nuclear power plants by fuel-coolant interactions when the external reactor vessel cooling strategy is failed. This phenomenon can cause shock wave that endangers surrounding reactor cavity wall due to resulting dynamic effects. Even though extensive researches have been performed to predict influences of the steam explosion, due to complexity of physical phenomena and thermal-hydraulic conditions, it is remained as one of possible hazards. The object of this study is to examine characteristics of reactor cavity and nuclear components under representative steam explosion conditions. In this context, an assembled finite element mesh was generated and evaluated by the trinitrotoluene model. As a result, stresses, strains and displacements of the reactor cavity and nuclear components were calculated. Subsequently, crack evaluation of reinforced concrete was performed and their results were discussed.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility

2012 ◽  
Author(s):  
Roberta Ferri ◽  
Fulvio Mascari ◽  
Paride Meloni ◽  
Giuseppe Vella
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Heat Removal ◽  
Nuclear Fission ◽  
Dynamic Coupling ◽  
Hydraulic Conditions ◽  
Loss Of Coolant ◽  
Accident Condition

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants. The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA. The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containment system dynamic coupling to cope with loss of coolant accidents. Full and complete nodalizations of SPES3 were developed for TRACE and RELAP5 codes in order to investigate the code response to the simulation of the same accidental transient. The DVI line DEG break was simulated in beyond design conditions, assuming the failure of all emergency heat removal systems and relying on PCC intervention for containment depressurization and decay heat removal. The comparison of the code simulation results, other than providing information on the system behavior, allowed to investigate specific phenomena evidenced by the codes, according to the related modeling approach of components with one and three-dimensional volumes. The TRACE and RELAP5 codes will be applied for further transient analyses and will be validated on SPES3 experimental data, once the facility will be available.

Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants

2004 ◽  
Author(s):  
Dilip Bhavnani ◽  
James Annett
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Supply System ◽  
Reactor Vessel ◽  
Original Design ◽  
Pressurized Water ◽  
Class 1 ◽  
Current Design ◽  
Successful Design

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.

The Application of an Integrated Head Assembly for Advanced Power Reactor 1400

2016 ◽  
Author(s):  
Tae Kyo Kang ◽  
Won Ho Jo ◽  
Yeon Ho Cho ◽  
Sang Gyoon Chang ◽  
Dae Hee Lee
Keyword(s):  
United Arab Emirates ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Power Reactor ◽  
Reactor Vessel ◽  
Head Region ◽  
The Usa ◽  
Head Assembly ◽  
Storage Area

The reactor vessel head region consists of a number of components and systems including reactor vessel head, CEDMs with their cables, cooling air system with ducts and fans, missile shield, seismic supports, head lift rig and cable supports. Prior to refueling operation, those components must be dismantled separately, and moved to the designated storage area. It was a very complicated and time consuming process. As a result, the integrated head assembly (IHA) was introduced to simplify those disassembling procedures, reduce refueling outage period, and improve safety in the containment building as those components are combined into a single system. To reduce refueling outage duration and radiation exposures to the workers by integrating the complicated reactor head region structures, KEPCO E&C has developed the IHA concept in the Korean Next Generation Reactor (KNGR) project [1]. The first application was implemented for the Optimized Power Reactor 1000 (OPR1000) at Shin-Kori units 1&2 and Shin-Wolsong units 1&2. With the past experience, the IHA was upgraded to be applied to the Advanced Power Reactor 1400 (APR1400). The design was patented in Korea [2], China, EU and the USA as modular reactor head area assembly. The IHA was applied for APR1400 nuclear power plants at Shin-Kori and Shin-Hanul, Korea. The design was also supplied to Barakah Nuclear Power Plants in the United Arab Emirates. This paper presents the design features and a variety of analysis which have been used for the APR1400 IHA.

scholarly journals Моделирование окисления расплава активной зоны ядерного реактора при наличии оксидной корки на поверхности расплава

2021 ◽  
Vol 91 (2) ◽  
pp. 232
Author(s):  
В.Б. Хабенский ◽  
В.И. Альмяшев ◽  
В.С. Грановский ◽  
Е.В. Крушинов ◽  
С.А. Витоль ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Solid Phase ◽  
Reactor Vessel ◽  
Hydrogen Release ◽  
Severe Accident ◽  
Water Steam ◽  
Exothermic Reactions ◽  
Realistic Situation ◽  
Water Reactors

At a severe accident of nuclear power plants with light-water reactors, the most effective way to localize the forming melt (corium) is to keep it in the cooled reactor vessel, the integrity of which depends on the value of heat flux from the melt to the reactor vessel. In this case, one of the critical processes is the melt oxidation by a water steam or a steam-air mixture. It process can lead to a significant increase in the thermal load on the reactor vessel due to a heat of exothermic reactions of oxidation of reducing agents, which presents in the melt, a thickness decreasing of the metallic part of the molten pool, and a hydrogen release. All of these factors strictly depends on the rate of oxidation. When considering the conditions of melt oxidation, it taken into account that for the accepted scenarios of a severe accident, the most realistic situation is the presence of a solid-phase oxide layer (oxidic crust) on the melt surface. Under these conditions, a dependence for calculating the rate of core melt oxidation based on the diffusion model proposed and its validation by using the obtained experimental data performed.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

Korean Regulatory Experience on PWSCC in Korean NPP

2010 ◽  
Author(s):  
Kyung-Cho Kim ◽  
Sung-bu Choi ◽  
Koo-Kab Chung ◽  
Hae-Dong Chung
Keyword(s):  
Residual Stress ◽  
Boric Acid ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Alloy 600 ◽  
Root Cause ◽  
Weld Material ◽  
Alloy 690

The degradation of alloy 600 and its weld material (alloy 82/182) has been reported in many nuclear power plants. In Korea, the crack induced by PWSCC was discovered in the drain nozzle of Yongkwang units 3 & 4 in 2006∼2008 and SG plug weld of Yongkwang unit 3 in 2007. In July 2007, during visual inspections of SG tube plugs at Yonggwang unit 4, boric acid deposits were observed around five Alloy 600 welded plugs. The root cause of the cracking in alloy 600 plugs was revealed to be due to the fact that the cracks were mainly caused by residual stress induced from the welding, expanding and tight-fitting. Younggwang unit 3 found the white small deposits on the drain nozzle on the 10th RFO in 2007. The root cause of the cracking in drain nozzle was revealed to be due to the initiation of a crack on the inside surface of drain nozzle and propagated to through wall cracks in the axial and circumferential direction. Younggwang unit 3 found the white widespread deposits on the upper head of a reactor vessel on the 12th RFO in 2010. Utility is trying to reveal the root cause of the cracking in the vent line of the reactor head according the KINS requirement. In this article, Korean regulatory experiences for PWSCC are introduced. After these PWSCC experiences, all SG tubes welded by Alloy 600 were replaced and all SG drain and instrumentation nozzles with Alloy 600 have been replaced into Alloy 690 material.

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