Methodology for Pressurized Thermal Shock Analysis in Nuclear Power Plant

摘要

The relevance of the fracture mechanics in the technology of the nuclear power plant isudmainly connected to the risk of a catastrophic brittle rupture of the reactor pressure vessel.udThere are no feasible countermeasures that can mitigate the effects of such an event thatudimpair the capability to maintain the core covered even in the case of properly functioningudof the emergency systems.udThe origin of the problem is related to the aggressive environment in which the vesseludoperates for long term (e.g. more than 40 years), characterized by high neutron flux duringudnormal operation. Over time, the vessel steel becomes progressively more brittle in theudregion adjacent to the core. If a vessel had a preexisting flaw of critical size and certainudsevere system transients occurred, this flaw could propagate rapidly through the vessel,udresulting in a through-wall crack. The severe transients that can lead the nuclear powerudplant in such conditions, known as Pressurized Thermal Shock (PTS), are characterized byudrapid cooling (i.e., thermal shock) of the a part of the internal reactor pressure vessel surfaceudthat may be combined with repressurization can create locally a sudden increase of theudstresses inside the vessel wall and lead to the suddenly growth of the flaw inside the vesseludthickness.udBased on the long operational experience from nuclear power plants equipped with reactorudpressure vessel all over the world, it is possible to conclude that the simultaneousudoccurrence of critical-size flaws, embrittled vessel, and a severe PTS transient is a very lowudprobability event. Moreover, additional studies performed at utilities and regulatoryudauthorities levels have shown that the RPV can operate well beyond the original design lifeud(40 years) because of the large safety margin adopted in the design phase.udA better understanding and knowledge of the materials behavior, improvement inudsimulating 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 thatudthe analysis performed during the 80’s were overly conservative, based on the tools andudknowledge available at that time.udNowadays the use of best estimate approach in the analyses, combined with tools for theuduncertainty evaluation is taking more consideration to reduce the safety margins, even fromudthe regulatory point of view. The US NRC has started the process to revise the technical baseudof the PTS analysis for a more risk-informed oriented approach. This change has the aim toudremove the un-quantified conservatisms in all the steps of the PTS analysis, from theudselection of the transients, the adopted codes and the criteria for conducting the analysisuditself thus allow a more realistic prediction.udThis change will not affect the safety, because beside the operational experience, severaludanalysis performed by thermal hydraulic, fracture mechanics and Probabilistic SafetyudAssessment (PSA) point of view, have shown that the reactor fleet has little probability ofudexceeding the limits on the frequency of reactor vessel failure established from NRCudguidelines on core damage frequency and large early release frequency through the periodudof license extension. These calculations demonstrate that, even through the period of licenseudextension, the likelihood of vessel failure attributable to PTS is extremely low (≈10-8/year)udfor all domestic pressurized water reactors.udDifferent analytical approaches have been developed for the evaluation of the safety marginudfor the brittle crack propagation in the rector pressure vessel under PTS conditions. Due toudthe different disciplines involved in the analysis: thermal-hydraulics, structural mechanicsudand fracture mechanics, different specialized computer codes are adopted for solving singleudpart of the problem.udThe aims of this chapter is to present all the steps of a typical PTS analysis base on theudmethodology developed at University of Pisa with discussion and example calculationudresults for each tool adopted and their use, based on a more realistic best estimate approach.udThis methodology starts with the analysis of the selected scenario by mean a SystemudThermal-Hydraulic (SYS-TH) code such as RELAP5 [2][3], RELAP5-3D [1], CATHARE2ud[4][6], etc. for the analysis of the global behavior of the plant and for the evaluation of theudprimary side pressure and fluid temperature at the down-comer inlet.udFor a more deep investigation of the cooling load on the rector pressure vessel internaludsurface at small scale, a Computational Fluid Dynamics (CFD) code is used. The calculatedudtemperature profile in the down-comer region is transferred to a Finite Element (FE)udstructural mechanics code for the evaluation of the stresses inside the RPV wall. The stressesudinduced by the pressure in the primary side are also evaluated.udThe stress intensity factor at crack tip is evaluated by mean the weight function methodudbased on a simple integration of the stresses along the crack border multiplied by the weightudfunction. The values obtained are compared with the critical stress intensity factor typical ofudthe reactor pressure vessel base material for the evaluation of the safety margin.