Results of benchmark calculations based on OLHF-1 test

Nicolas, L.; Durin, M.; Koundy, V.; Mathet, Eric; Andrea, Bucalossi; Eisert, Peter; Sievers, J.; Humphries, Larry L.; Smith, Jack; Pistora, Vladislav; Ikonen, Kari

doi:10.1016/s0029-5493(03)00064-5

Cited by 26 publications

(3 citation statements)

References 1 publication

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The results of benchmark calculations in the OLHF project can be summarized as follows [65]: l Generally, the predicted failure times calculated by participants agree reasonably well with the test data. All the models, regardless of their complexity, gave reasonable prediction of the failure time (within 15 minutes).…”

Section: Scaled Experiments On Vessel Failurementioning

confidence: 65%

In-Vessel Core Degradation

2012

Nuclear Safety in Light Water Reactors

View full text Add to dashboard Cite

Section: Scaled Experiments On Vessel Failurementioning

confidence: 65%

In-Vessel Core Degradation

2012

Nuclear Safety in Light Water Reactors

View full text Add to dashboard Cite

“…We should note that, both experimentally and numerically, the failure mode is not clearly identified for the OLHF1 experiment. This difficulty was already underlined in the benchmark performed on this experiment with different calculation codes (Nicolas et al, 2003). However, in that case, the rupture times evaluated by ASTEC for the different failure modes are sufficiently close for the simulation of accident scenarios in a reactor.…”

Section: Fig14 -Olhf1 Experiments -Temperature Evolution At Vessel Ou...mentioning

confidence: 78%

Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

Carénini¹,

Fleurot²,

Fichot³

2014

Nuclear Engineering and Design

View full text Add to dashboard Cite

“…All three programs provided data for model development and validation, and have led to a better understanding of the mechanical behavior of the RPV lower head (Koundy and Hoang, 2008). The numerical analyses revealed that there are still uncertainties and deviations between analytical codes and models (Nicolas et al, 2003;Willschütz et al, 2006). In order to improve predictions, more investigation and work are needed on the choice of the failure criteria and failure mode.…”

Section: Introductionmentioning

confidence: 99%

A thermal structural analysis tool for RPV lower head behavior during severe accidents with core melt

Madokoro

Miassoedov

Schulenberg

2018

Mechanical Engineering Letters

View full text Add to dashboard Cite

The Fukushima Daiichi Nuclear Power Plant accident showed a significant effect on the environment due to the release of a large amount of radioactive material. To prevent the damage of the reactor pressure vessel (RPV), an in-vessel melt retention (IVR) by external reactor vessel cooling is considered to be a key severe accident management strategy. For its success, investigation of the thermal-mechanical behavior of the RPV lower head is of importance. The main objective of this study is the development of a thermal structural calculation tool for simulating the failure process and the visco-plastic behavior of the RPV lower head wall during the late phase of a core-melt severe accident. OpenFOAM, an open-source toolbox for developing numerical solvers, was used for the calculation platform. Thermal behavior of the molten pool in the RPV lower head is simulated by the phasechange effective convectivity model (PECM). One of the LIVE (Late In-Vessel Phase Experiments) experiments was analyzed for an evaluation purpose and reasonable results were obtained. The solver was then extended to couple PECM with a structural analysis model that considers thermal expansion, plasticity, creep and material damage. Two FOREVER (Failure Of Reactor Vessel Retention) tests, each of which uses different type of steel, were analyzed: EC-FOREVER-2 with French RPV steel 16MND5 and EC-FOREVER-4 with American RPV steel SA533B1. The deformation and failure time agreed reasonably well with the measured data. In addition, the failure mode was also well predicted qualitatively. The numerical analysis showed that the developed tool has a capability of simulating the lower head thermal structural behavior.

show abstract

Results of benchmark calculations based on OLHF-1 test

Cited by 26 publications

References 1 publication

In-Vessel Core Degradation

In-Vessel Core Degradation

Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

A thermal structural analysis tool for RPV lower head behavior during severe accidents with core melt

Contact Info

Product

Resources

About