Failure assessment of RPV nozzle under loss of coolant accident

Siegele, Dieter; Hodulak, L.; Varfolomeyev, Igor; Nagel, Gerhard

doi:10.1016/s0029-5493(99)00184-3

Cited by 15 publications

(4 citation statements)

References 1 publication

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In 1999, Siegele et al [10], performed fracture mechanics analysis of the RPV's nozzle under loss of coolant accident (LOCA). In the study, they have presented that the nozzle regions face higher stresses and lower temperature in the event of the LOCA, and is therefore vulnerable and may threaten the integrity of the RPV.…”

Section: Literature Reviewmentioning

confidence: 99%

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Murtaza

Hyder

2018

Engineering Fracture Mechanics

View full text Add to dashboard Cite

Section: Literature Reviewmentioning

confidence: 99%

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Murtaza

Hyder

2018

Engineering Fracture Mechanics

View full text Add to dashboard Cite

“…The severe cooling of the core followed by repressurization poses a serious challenge to the structural integrity of RPV. It should be noted that the inlet nozzle is a weak link of RPV, where the initiation and propagation of crack are easy to be produced in the thermo‐mechanical field. On the other hand, the inner surface of the vessel is exposed to intense neutron irradiation, and the fracture toughness of material is greatly reduced .…”

Section: Introductionmentioning

confidence: 99%

“…The severe cooling of the core followed by repressurization poses a serious challenge to the Nomenclature: E, elastic modulus; R i , inner radius; p, internal pressure; T, temperature; h, heat transfer coefficient; t, thickness of the base wall; t clad , thickness of the cladding; K c , fracture toughness; RT NDT , nil-ductility reference temperature; M, average safety margin; ΔRT NDT , shift of reference temperature; G c , critical energy release rate structural integrity of RPV. It should be noted that the inlet nozzle is a weak link of RPV, 1 where the initiation and propagation of crack are easy to be produced in the thermo-mechanical field. On the other hand, the inner surface of the vessel is exposed to intense neutron irradiation, and the fracture toughness of material is greatly reduced.…”

Section: Introductionmentioning

confidence: 99%

Nonlinear numerical study of crack initiation and propagation in a reactor pressure vessel under pressurized thermal shock using XFEM

Sun

Chai

Bao

2017

Fatigue Fract Eng Mat Struct

View full text Add to dashboard Cite

Under pressurized thermal shock (PTS), once crack initiation occurs in a reactor pressure vessel (RPV), the stress concentration around the crack tips may result in local instability and crack propagation. The temperature‐dependent material properties are introduced into the finite element analysis model. According to the response of the transient temperature field and stress field near the crack tip region, the influence of PTS on the carrying capacity of the structure is demonstrated. Also, the process of crack initiation and propagation is simulated by using the extended finite element method (XFEM). The results show that the crack of mode I is easy to be initiated on the nozzle in the initial high temperature and high pressure working state. The effect of cladding on the RPV integrity is enhanced with the increase of crack size, and it mainly depends on the crack length. During the PTS transient, a discontinuity of stress exists near the cladding‐base interface, and the cladding bears more loading than the base of the same size. As the unloading of the internal pressure, the stress does not decline due to the strong thermal shock. In the late stage of the PTS process, the internal pressure caused by repressurization poses a challenge to the strength of the structure. With the decrease of the base wall thickness, the allowable repressure value also gradually decreases. However, the increase of the base wall thickness causes the rise of thermal stress, and the allowable repressure load is not significantly improved.

show abstract

“…In general, the most accurate method of analysis is to use elastic-plastic Finite Element Analysis (FEA) of the cracked component, and calculate the J-integral as a function of position on the crack tip line [3]. Many examples of crackedbody FEA applied to PTS events in RPVs exist, including examples of both elastic analysis [15] and inelastic analysis [16], [17]. The elastic-plastic cracked-body FEA approach is used in the current study.…”

Section: Introductionmentioning

confidence: 99%

Parametric design of scaled-down pressurised thermal shock test specimens using inelastic analysis

Coules

Orrock²,

Truman³

2017

Engineering Fracture Mechanics

View full text Add to dashboard Cite

General rightsThis document is made available in accordance with publisher policies. Please cite only the published version using the reference above. AbstractPressurised thermal shock tests that simulate fracture conditions relevant to nuclear reactor pressure vessels are difficult to perform and require very large specimens. This study examines the feasibility of scaled-down specimens to allow easier testing. A series of finite element models was used to design scaled-down specimens that produce a crack-driving force, crack tip temperature trajectory and constraint conditions very similar to those which occurred in a previous large-scale spinning cylinder experiment known as NESC-1. It is shown that equivalent conditions can be achieved in much smaller specimens than NESC-1 using a practical set of testing parameters.Keywords: Thermal shock, finite element analysis, pressure vessel, spinning cylinder, J-integral Graphical abstractHighlights  Thermal shock tests for nuclear reactor pressure vessels can be miniaturised.  Models show that scaled-down tests can give equivalent fracture initiation conditions.  Specimens equivalent to the NESC-1 thermal shock test have been designed.

show abstract

Failure assessment of RPV nozzle under loss of coolant accident

Cited by 15 publications

References 1 publication

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Nonlinear numerical study of crack initiation and propagation in a reactor pressure vessel under pressurized thermal shock using XFEM

Parametric design of scaled-down pressurised thermal shock test specimens using inelastic analysis

Contact Info

Product

Resources

About