Structural Integrity of a Reactor Vessel Lower Head Under In-Vessel Steam Explosion Loads

2013 ◽  
Author(s):  
Guohong Xue ◽  
Yinbiao He ◽  
Ming Cao ◽  
Hao Yu ◽  
Yongjian Gao
Keyword(s):  
Steam Explosion ◽  
Power Plants ◽  
Structural Integrity ◽  
Equivalent Plastic Strain ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Conservation Of Energy ◽  
Lower Head ◽  
Pressure Pulses

Passive nuclear power plants emphasize the “In vessel retention” idea such that, after a postulated severe accident event, the reactor vessel wall, flooded with emergency cooling water, will maintain its structural integrity and consequently keep the molten core inside the reactor vessel. However, steam explosion may still occur when the melting core or molten metal is mixed with cooling water. The huge pressure pulses from the steam explosion may be a threat to the structural integrity of the reactor vessel lower head and the potential failure may make the situation difficult to control. This paper presents a detailed analysis on the structural integrity of a reactor vessel lower head. First, a mathematical model is built to relate the equivalent plastic strain in the lower head under explosive loads based on the law of conservation of energy. Then a finite element model, using the computer code ABAQUS, is built and the material’s yield strength as a function of strain rate was simulated using the Bodner-Symonds methodology. With this model, the dynamic response and the structural integrity of the reactor vessel lower head is studied, considering the effect of the magnitude, the shape and the duration of the pressure pulses. The method used in this paper is believed to be applicable to other types of devices containing potential explosive materials and thus could provide guiding significance to similar problems.

Thermal and Structural Analysis of Reactor Vessel Lower Head Considering Core Meltdown Accident

2018 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun ◽  
Peng Tang
Keyword(s):  
Structural Analysis ◽  
Creep Deformation ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Creep Failure ◽  
The Core ◽  
Lower Head ◽  
Core Meltdown

In the core meltdown severe accident, in-vessel retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) is an important mitigation strategy. During the IVR strategy, the core debris forming a melt pool in the reactor pressure vessel (RPV) lower head (LH) will produce extremely high thermal and mechanical loadings to the RPV, which may cause the failure of RPV due to over-deformation of plasticity or creep. Therefore, it is necessary to study the thermomechanical behavior of the reactor vessel LH during IVR condition. In this paper, under the assumption of IVR-ERVC, the thermal and structural analysis for the RPV lower head is completed by finite element method. The temperature field and stress field of the RPV wall, and the plastic deformation and creep deformation of the lower head are obtained by calculation. Plasticity and creep failure analysis is conducted as well. Results show that under the assumed conditions, the head will not fail due to excessive creep deformation within 200 hours. The results can provide basis for structural integrity analysis of pressure vessels.

scholarly journals Analysis of criticality of melt during severe accidents in reactor vessel

2018 ◽  
pp. 3-10
Author(s):  
Yu. Kovbasenko ◽  
Yevgen Bilodid
Keyword(s):  
Boric Acid ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Cooling Water ◽  
Nuclear Fission ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Emergency Situations ◽  
The Core

The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

scholarly journals Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Alejandro Nuñez-Carrera ◽  
Raúl Camargo-Camargo ◽  
Gilberto Espinosa-Paredes ◽  
Adrián López-García
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Detailed Model ◽  
Lower Head ◽  
Reactor Pressure

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

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.

Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
K. M. Koo ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Heat Removal ◽  
Elevated Pressure ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Severe Accident ◽  
Experimental Facility ◽  
Dual Strategy ◽  
Accident Management ◽  
Lower Head ◽  
Inner Diameter

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.

Analysis on Ex-Vessel Steam Explosion in CPR1000+ Unit

2010 ◽  
Author(s):  
Juanhua Zhang ◽  
Jiming Lin ◽  
Shishun Zhang
Keyword(s):  
Physical Model ◽  
Steam Explosion ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Cavity Wall ◽  
Severe Accident ◽  
Severe Accidents ◽  
Structural Capacity ◽  
Impulse Load

Reactor Pit Flooding System (RPF) is adopted under the severe accidents situation in CPR1000+ units. It can move the heat generated from the reactor core via external reactor vessel cooling (ERVC) to keep the integrity of RPV and achieve the in-vessel corium retention (IVR). But if IVR function of RPF is failed, there is Ex-Vessel Steam Explosion (EX-SE) risk. The Ex-Vessel Steam Explosion is analyzed by MC3D software which is for fuel and cooling interaction (FCI). The physical model of CPR1000+ for Steam Explosion is built firstly and then the phenomenon of Ex-Vessel Steam Explosion under typical severe accident is analyzed. The conclusion of this study is that the impulse load of pressure on the cavity wall induced by steam explosion is about 310KPas ∼ 440KPas. Referencing the structure capacity of AP600 containment, if the structural capacity of CPR1000+ containment is equal to AP600, the impulse load of pressure is lower than it. So it could be preliminarily estimated that steam explosion will not threaten the integrality of CPR1000+ containment.

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].

In-Vessel Corium Behavior and Steam Explosion Risks: A Review of Experimental Approaches

2017 ◽  
Author(s):  
Pei Shen ◽  
Wenzhong Zhou
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Radioactive Material ◽  
Severe Accident ◽  
Reactor Accident ◽  
Experimental Approaches

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

Creep Evaluation for a PWR Reactor Pressure Vessel Lower Head Under Severe Accident Conditions Considering Sustained Internal Pressure

2018 ◽  
Author(s):  
YongJian Gao ◽  
Ming Cao ◽  
YinBiao He
Keyword(s):  
Internal Pressure ◽  
Failure Modes ◽  
Structural Integrity ◽  
Creep Damage ◽  
Pressure Vessels ◽  
Mitigation Strategies ◽  
Field Analysis ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of appropriate severe accident mitigation strategies for AP1000 Nuclear Power Plant (NPP), and assurance of prevention against to thermal failure and structural failure of Reactor Pressure Vessels (RPV) is the prerequisite of IVR. A Finite Element Model fora RPV considering lower head melting was established, the creep calculation was carried out after the temperature field analysis, and the stress-strain responses for different times were obtained. By means of choosing representative evaluation sections and applying the Accumulative Damage Theory based on Larson-Miller Parameter, the Creep Damage calculations and evaluations were conducted. The results showed that the failure modes associated with creep rupture would not happen under IVR condition when a certain amount of internal pressure sustained. The approaches employed in this paper could be utilized in structural integrity evaluation of RPV under IVR for other new type NPPs.

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