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Application of REPAS methodology to assess the reliability of passive safety systems
No full textThe paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS) methodology developed by
University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system
in order to increase the confidence toward its operation and to compare the performances of active and passive systems and
the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their
goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment
studies and it can be seen as a support to Probabilistic Safety Analysis studies.With regard to this, some examples in the application
of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the
analyses and tomodel the selected systems. Probability distributions have been assigned to the uncertain input parameters through
engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks’ formula has been taken into
account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets
Use of Natural Circulation Flow Map
No full textThis lecture describes the different loop flow regimes encountered in single-phase and two-phase natural circulation. The database gathered from ten integral system experiments has been used to establish a Natural Circulation Flow Map that can be used to characterize the effectiveness of a natural circulation system
Fracture Mechanics Analysis in a Pressurized Heavy Water Reactor Vessel during LOCA Scenario
No full textThe Reactor Pressure Vessel (RPV) has long been considered one of the most reliable components in Pressurized Water Reactor (PWR). Nowadays, a general target for the countries that produce nuclear energy is to extend the operation life of existing plants. From this point of view, the RPV is one of the major components that may limit the useful life of the nuclear plant. The risk for the RPV structural integrity is connected to the presence of a flaw of sufficient size, a high level of embrittlement due to radiation damage, and the occurrence of a thermal-hydraulic transient inducing strong stresses in the vessel wall.
Severe loading conditions are produced during a Pressurized Thermal Shock (PTS) event, in which an overcooling may induce strong thermal stresses while the internal pressure can be maintained at high level or the system can be re-pressurized during the transient. Such conditions are generated in a Loss Of Coolant Accident (LOCA) transient during the emergency injection.
In recent years, important progresses have been made in the development of analysis methods and tools for the best estimation of the thermal and pressure loads on the vessel wall. In this direction, the US-NRC published in 2007 a document aimed at reviewing the rules adopted in PTS analysis, established in the 1980s, containing significant conservatisms, for a Best Estimate (BE) approach combined with uncertainty assessment.
In this paper, the methodology for Fracture Mechanics analysis developed at University of Pisa aimed to perform parametric analysis assuming various shapes and locations of the flaw is applied to a Pressurized Heavy Water Reactors (PHWR) during a LOCA scenario. Four steps can be identified starting from the thermal hydraulic analysis of the Nuclear Power Plant (NPP) behaviour with Relap5-3D© in order to calculate the cooling loads of the Emergency Core Coolant Systems (ECCS). The second step is the analysis by mean a CFD code (CFX) of the mixing phenomena occurring in the Down-Comer (DC) and the calculation of the thermal load on the RPV internal surface. The third step is represented by the evaluation of the stresses inside the RPV wall by mean a Finite Element (FE) code (Ansys) under the thermal and pressure loads calculate in the previous steps. The last step is represented by the calculation of the Stress Intensity Factor (SIF) KI by mean the Weight Function method and the comparison with the critical SIF KIc of the material, once the stresses inside the undamaged RPV wall are known.
The goal of this work is the evaluation of the safety margin for the operation of the RPV, adopting a BE approach in all the steps of the analysis. This result will be compared with the one obtained with the application of the ASME XI criteria for the KI evaluation with the aim to show that the BE approach leads to a larger safety margin
Methodology for Pressurized Thermal Shock Analysis in Nuclear Power Plant
Full text linkThe relevance of the fracture mechanics in the technology of the nuclear power plant is
mainly connected to the risk of a catastrophic brittle rupture of the reactor pressure vessel.
There are no feasible countermeasures that can mitigate the effects of such an event that
impair the capability to maintain the core covered even in the case of properly functioning
of the emergency systems.
The origin of the problem is related to the aggressive environment in which the vessel
operates for long term (e.g. more than 40 years), characterized by high neutron flux during
normal operation. Over time, the vessel steel becomes progressively more brittle in the
region adjacent to the core. If a vessel had a preexisting flaw of critical size and certain
severe system transients occurred, this flaw could propagate rapidly through the vessel,
resulting in a through-wall crack. The severe transients that can lead the nuclear power
plant in such conditions, known as Pressurized Thermal Shock (PTS), are characterized by
rapid cooling (i.e., thermal shock) of the a part of the internal reactor pressure vessel surface
that may be combined with repressurization can create locally a sudden increase of the
stresses inside the vessel wall and lead to the suddenly growth of the flaw inside the vessel
thickness.
Based on the long operational experience from nuclear power plants equipped with reactor
pressure vessel all over the world, it is possible to conclude that the simultaneous
occurrence of critical-size flaws, embrittled vessel, and a severe PTS transient is a very low
probability event. Moreover, additional studies performed at utilities and regulatory
authorities levels have shown that the RPV can operate well beyond the original design life
(40 years) because of the large safety margin adopted in the design phase.
A better understanding and knowledge of the materials behavior, improvement in
simulating in a more realistic way the plant systems and operational characteristics and a better evaluation of the loads on the RPV wall during the PTS scenarios, have shown that
the analysis performed during the 80’s were overly conservative, based on the tools and
knowledge available at that time.
Nowadays the use of best estimate approach in the analyses, combined with tools for the
uncertainty evaluation is taking more consideration to reduce the safety margins, even from
the regulatory point of view. The US NRC has started the process to revise the technical base
of the PTS analysis for a more risk-informed oriented approach. This change has the aim to
remove the un-quantified conservatisms in all the steps of the PTS analysis, from the
selection of the transients, the adopted codes and the criteria for conducting the analysis
itself thus allow a more realistic prediction.
This change will not affect the safety, because beside the operational experience, several
analysis performed by thermal hydraulic, fracture mechanics and Probabilistic Safety
Assessment (PSA) point of view, have shown that the reactor fleet has little probability of
exceeding the limits on the frequency of reactor vessel failure established from NRC
guidelines on core damage frequency and large early release frequency through the period
of license extension. These calculations demonstrate that, even through the period of license
extension, the likelihood of vessel failure attributable to PTS is extremely low (≈10-8/year)
for all domestic pressurized water reactors.
Different analytical approaches have been developed for the evaluation of the safety margin
for the brittle crack propagation in the rector pressure vessel under PTS conditions. Due to
the different disciplines involved in the analysis: thermal-hydraulics, structural mechanics
and fracture mechanics, different specialized computer codes are adopted for solving single
part of the problem.
The aims of this chapter is to present all the steps of a typical PTS analysis base on the
methodology developed at University of Pisa with discussion and example calculation
results for each tool adopted and their use, based on a more realistic best estimate approach.
This methodology starts with the analysis of the selected scenario by mean a System
Thermal-Hydraulic (SYS-TH) code such as RELAP5 [2][3], RELAP5-3D [1], CATHARE2
[4][6], etc. for the analysis of the global behavior of the plant and for the evaluation of the
primary side pressure and fluid temperature at the down-comer inlet.
For a more deep investigation of the cooling load on the rector pressure vessel internal
surface at small scale, a Computational Fluid Dynamics (CFD) code is used. The calculated
temperature profile in the down-comer region is transferred to a Finite Element (FE)
structural mechanics code for the evaluation of the stresses inside the RPV wall. The stresses
induced by the pressure in the primary side are also evaluated.
The stress intensity factor at crack tip is evaluated by mean the weight function method
based on a simple integration of the stresses along the crack border multiplied by the weight
function. The values obtained are compared with the critical stress intensity factor typical of
the reactor pressure vessel base material for the evaluation of the safety margin
Systematic Comparison of Horizontal and Vertical Steam Generator Performance during Long Term Station Black Out Conditions using Relap5 and Cathare2
No full textStation black out conditions (i.e. loss of off-site power, with simultaneous failure of all diesel generators)
over a period of several hours are a very unlikely, but can develop severe consequences. Previous
analysis has shown that a well planned emergency operating procedure, utilizing just equipment powered
by batteries, can considerably extend the grace period, and mitigate severe accident progression.
The current paper presents an analysis with Relap5 and Cathare2 system codes of the heat transfer
capabilities of horizontal and vertical steam generators during the progression of a station black out
transient. A long term station black out transient is characterized by high pressure in the primary side, and
removal of the decay heat via natural circulation by the inventory present in the steam generators at the
time of the initiating event. Secondary side mass inventory boils down in a certain period of time,
depending on the reactor type. Subject of the presented analysis are the dependence of the heat transfer
on the level of the steam generator, the different behaviour of horizontal and vertical steam generators in
this respect, and the consequences for main primary side parameters such as average coolant
temperature and primary pressure.
The analysis shows that in natural circulation regime only few centimetres of the U-tubes are needed to
cool the primary fluid to the temperature of the secondary side. Vertical steam generators are cooling the
primary fluid down to the temperature of the secondary side until the level of the steam generator stays
above the U-tubes cooling length. Horizontal steam generators on the other hand show an increasing
number of U-tubes completely uncovered with decreasing level. As a result the primary side coolant
temperature at the SG outlet is continuously increasing with decreasing SG level. The consequence is a
different primary side pressure and temperature trend during a SBO transient in NPP with horizontal and
vertical steam generators
Fracture Mechanics Analysis of Generic WWER-1000 RPV in PTS Event
No full textThe considered research activity deals with the application of a chain of numerical codes, in order to develop a computational tool for Pressurized Thermal Shock (PTS) analysis. Nuclear reactor pressure vessel structural integrity is concerned regarding the risk of brittle fracture. Special reference is made to the identification of the 3D Stress-Strain State of the WWER-1000 pressure vessel. A Main Steam Line Break (MSLB) accident has been taken as reference scenario for the study. The thermal hydraulic calculation has been carried out using Relap5/mod3.3 code. The results are used for computing the combined effect of thermal and pressure loading, which constitutes the initial event for the PTS. Those data supply the input for the structural mechanics code. The Stress-Strain State of the WWER-1000 reactor pressure vessel has been evaluated by means of the Ansys 5.7 code
Effect of SG modeling on primary side pressure and temperature trend during long term station black out conditions in VVER using Relap5 and Cathare2 system
No full textAlthough station black out conditions (i.e. loss of off-site power with simultaneous failure of all diesel generators) over a period of several hours are very unlikely, analysis of station black out transients and possible countermeasures are important. A well planned emergency operating procedure, utilizing just equipment powered by batteries, can extend the grace period considerably, and mitigate severe accident progression. Until the occurrence of core damage best estimate thermal hydraulic system codes like Relap5 or Cathare2 can give a good picture of the development of the transient.
Heat transfer from the primary to the secondary side is one of the dominant phenomena during the first hours of a station black out transient. Primary side pressure stays high, and the decay heat is transferred by natural circulation to the secondary side. The steam generator inventory boils off, and the steam generator level decreases. Modelling of the steam generator therefore is of special importance. The current paper presents an analysis for VVER-1000 type steam generators (horizontal) with different nodalisation detail and different codes (Cathare2 and Relap5). The nodalisation approach has been verified by comparing calculated results to PSB-VVER steam generator experiments.
The results show that codes like Relap5 and Cathare2, by assuming uniform bulk flow properties over the flow cross section and within volume units, introduce a distortion compared to the real physics in case station black out conditions. The reason lies in the specifics of the horizontal steam generator. While the heat transfer surface decreases continuously in the actual steam generator, the necessity of modelling the steam generator with nodes of finite volume creates a step-wise reduction of the heat transfer surface. The effects on the primary side temperature and pressure trends are not negligible, and decrease with increasing detail of modelling
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