Study on Ultimate Load Capacity of Reactor Pressure Vessel under Critical Heat Flux

2017 ◽  
Vol 53 (2) ◽  
pp. 45 ◽  
Author(s):  
Jianwei ZHU
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Ultimate Load ◽  
Load Capacity ◽  
Reactor Pressure Vessel ◽  
Ultimate Load Capacity ◽  
Reactor Pressure

Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure

2016 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianfeng Mao ◽  
Jianwei Zhu ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Heat Flux ◽  
Internal Pressure ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Wide Range ◽  
Depth Analysis ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

Enhancement of RPV Critical Heat Flux by Injection Flow

2017 ◽  
Author(s):  
Wei Chen ◽  
Canhui Sun ◽  
Jun Geng
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Metallic Layer ◽  
Fluid Temperature ◽  
Injection Flow ◽  
Severe Accidents ◽  
Reactor Pressure ◽  
Residual Heat

Under severe accidents, the reactor pressure vessel is flooded with water and the residual heat is removed by two-phase natural circulation through the flow channel between the reactor vessel and thermal insulation. If the heat flux of the outer wall is lower than local critical heat flux, the residual heat can be removed, and if the heat flux of the outer wall is higher than local critical heat flux, the reactor pressure vessel should be molten. For AP-type reactors, like AP1000 and CAP1400, critical heat flux of the reactor pressure vessel is the heat transfer limits of residual heat under severe accidents. Previous studies indicate that after severe accident a two-layer molten pool can be formed, namely metallic layer and oxide layer. Compared with oxide layer, in metallic layer, the heat flux more easily exceeds the heat transfer limits due to its low thermal resistance. In this study, an approach was proposed to enhance local critical heat flux. This approach is expected to be used in local area around reactor pressure wall, like metallic layer, to increase the reactor pressure vessel intact probability under severe accidents. In this new approach, injection flow channels are added to the present flow channel by adding simple flow pipes from insulation near 60 to 80 degree where exceeding critical heat flux is most likely to happen. The fluid flow under external reactor vessel cooling (ERVC) condition is divided into two parts: one part is from downward (0 degree) to upward (90 degree) along the curved reactor pressure vessel and the other part is from injection pipe (about 70 degree) to upward. The fluid temperature from injection pipe is lower than that from downward due to residual heat from the reactor pressure wall. And hence, the local critical heat flux is likely to increase because of inject turbulence and low fluid temperature. An experimental facility is conducted to study the mechanism of injection influence on critical heat flux under normal pressure condition. There are two main loops in this facility: one is main loop while the other is injection loop. The test section is an inclined downward heated rectangular channel with its inclined angle varied from 0 degree to 90 degree. Flow and thermal conditions are listed: in main loop, mass flow velocity ranging from 100kg/m2s to 600kg/m2s with fluid temperature from 90 °C to 105 °C; In injection loop, mass flow velocity ranging from 0 to 600 kg/m2s with fluid temperature from 85 °C to 105 °C. Under the above condition, with and without injection flow, critical heat flux experiments were conducted. It indicates that injection velocity has great effect on critical heat flux, while injection subcooled has little effect. The critical heat flux can be increased by 0.07MW/m2 to 0.33MW/m2 depending on various injection velocities and main loop conditions.

scholarly journals Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-14 ◽  
Author(s):  
Mindaugas Valinčius ◽  
Tadas Kaliatka ◽  
Algirdas Kaliatka ◽  
Eugenijus Ušpuras
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Pressure Vessel ◽  
Heat Fluxes ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Focusing Effect ◽  
Accident Management ◽  
Lower Head ◽  
Reactor Pressure

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

Prediction of Subcooled Water Flow Boiling Critical Heat Flux (CHF) at Low Pressure

2017 ◽  
Author(s):  
Fangxin Hou ◽  
Huajian Chang ◽  
Yufeng Zhao ◽  
Ming Zhang ◽  
Peipei Chen ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Water Flow ◽  
Pressure Vessel ◽  
High Speed ◽  
Flow Boiling ◽  
Reactor Vessel ◽  
Lower Head ◽  
Inclination Angles ◽  
Reactor Pressure

In vessel retention (IVR) is one of the key severe accident mitigation strategies to maintain reactor pressure vessel (RPV) integrity. IVR designs utilize the reactor pressure vessel lower head to contain molten fuel and rely on external reactor vessel cooling (ERVC) to remove decay heat. The capacity of ERVC is limited by the critical heat flux (CHF) of flow boiling on the outside of the reactor vessel surface. Therefore, the determination of critical heat flux (CHF) is crucial to predict whether the adoption of IVR would be successful in mitigating severe accidents. In 1999, Celeta et.al proposed a superheated layer vapor replenishment model. In this model they postulated that CHF would occur when the superheated layer was occupied by the vapor blanket coming into contact with the heated wall and they successfully predicted the critical heat flux in subcooled water flow boiling under high mass flux, high liquid subcooling and low/medium pressure conditions. To evaluate the practicability of this model in predicting CHF under IVR conditions, CHF experiments were performed under natural circulation conditions on the experiment facility “Test of External Vessel Surface with Enhanced Cooling” (TESEC). Experiments are conducted in a 30 mm wide, 61mm high rectangular flow channel with a 200 mm long heated surface along the flow direction. Two quartz windows are installed at the sidewalls of the flow channel for visualization. In order to simulate various positions of the reactor lower head, experiments at different inclination angles of the test section were conducted. The high speed visualization data at CHF point at various inclination angles were processed and analyzed by a MATLAB code developed by the author. The vapor blanket thickness at various inclination angles was measured from the visualization data and was also predicted by the Celeta model. By using geometry data from high speed images, CHF values were calculated by Celeta model and compared with the experimental results at various inclination angles. Limitations of the Celeta model in adaptation of predicting CHF under IVR conditions were further discussed.

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure

The effects of thermal stress on the crack propagation in AP1000 reactor pressure vessel

2020 ◽  
Vol 110 ◽  
pp. 102798
Author(s):  
KaiTai Liu ◽  
Mei Huang ◽  
JunJie Lin ◽  
HaiPeng Jiang ◽  
BoXue Wang ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Crack Propagation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Ap1000 Reactor ◽  
Reactor Pressure

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

The ageing behavior and the correlation between hardness and coercivity of Cu-rich reactor pressure vessel model steels

2021 ◽  
Vol 527 ◽  
pp. 167698
Author(s):  
Xuejiao Wang ◽  
Wenjiang Qiang ◽  
Guogang Shu ◽  
Junwei Qiao ◽  
Yucheng Wu
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure
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