Scoping Analyses of Parameterized Cool-Down Transients Associated With Reactor Shutdown

The United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project in 1999 to determine if improved technologies can provide a technical basis to reduce the conservatism in the current regulations for pressurized thermal shock (PTS) while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license renewal considerations. During the PTS re-evaluation study, an improved risk-informed computational methodology was developed that provides a more realistic characterization of PTS risk. This updated methodology was recently applied to three commercial PWRs. The results of this study provide encouragement that a technical basis can be established to support a relaxation of current PTS regulations. One significant model improvement applied in the PTS re-evaluation study was the development of flaw databases derived from the non-destructive and destructive examinations of material from cancelled reactor pressure vessels (RPV). Empirically-based statistical distributions derived from these databases and expert illicitation were used to postulate the number, size, and location of flaws in welded and base metal (plate and forging) regions of an RPV during probabilistic fracture mechanics (PFM) analyses of RPVs subjected to transient loading conditions such as PTS. However, limitations in the available flaw data have required assumptions to be made to complete the risk-based flaw models. Sensitivity analyses were performed to evaluate the impact of four flaw-related assumptions. Analyses addressed: 1) truncations of distributions to exclude flaws of extreme depth dimensions, 2) vessel-to-vessel differences in flaw data, 3) large flaws observed in weld repair regions, and 4) the basis for estimating the number of surface breaking flaws. None of the four alternate weld flaw models significantly impacted calculated vessel failure frequencies or invalidated the tentative conclusions derived from the USNRC PTS re-evaluation study.

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life. Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a). This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.

To assess the structural integrity of reactor vessels (RVs) during pressurized thermal shock (PTS) events, a deterministic fracture mechanics (DFM) approach has been widely used such as the procedure in JEAC4206-2007. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RV has become attractive recently because uncertainties related to input parameters can be incorporated rationally. The probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. In this paper, in order to verify the applicability of a PFM method to JEAC4206-2007, deterministic and probabilistic analyses have been performed, and the effects of initial crack size defined in JEAC4206-2007 on the temperature margin obtained from DFM and the probability of crack initiation obtained from PFM have been evaluated. With regard to the PTS event variation, a stuck open valve scenario (SO) has been considered in addition to large- and small-break loss of coolant accident (LBLOCA, SBLOCA) and main steam line break (MSLB).

Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

Improvements to a probabilistic fracture mechanics code for evaluating the integrity of a RPV under transient loading

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

The structural components that comprise nuclear reactors and their supporting structures are subjected to harsh operating environments that can challenge their integrity, especially after exposure for extended durations or under accident condition. As one of the most significant components of a Reactor, the Reactor Pressure Vessel (RPV) is exposed to an aggressive environment during the operation time (e.g. more than 40 years). Ageing degradation mechanisms (e.g. thermos-fatigue) could grow initial defects up to a critical size, increasing the susceptibility to failure in the RPV. The conventional methods are mostly based on simple crack and structure geometries. Very limited studies consider the real conditions of the RPV subjected to a thermal shock due to a Loss of Coolant Accident (LOCA). During a LOCA event, the most severe conditions take place when the emergency core cooling (ECC) water is injected inside the cold legs filled initially with hotter water and/or steam. The rapid cooling of the down-comer and the internal RPV surface followed probably by re-pressurization of the RPV causes large temperature gradients and variation of pressure which induces thermal-mechanical stresses. In order to develop the model for integrity assessment of a reactor pressure vessel (RPV) subjected to pressurized thermal shock (PTS), a multi-physics simulation, which includes the thermo-hydraulic, thermo-mechanical and fracture mechanics analyses is necessary. The multi-physics simulations are performed using Cardinal, a wrapping of the GPU-oriented spectral element Computational Fluid Dynamics (CFD) code NekRS and other multi-physics sub-modules within the MOOSE framework. Cardinal now fully supports MOOSE stochastic perturbations of NekRS models with varying boundary conditions, initial conditions, material properties, and any other quantity which is defined by a kernel (such as coefficients in a momentum source model). The implementation is designed in a flexible manner so that scalar values are sent from MOOSE into a user scratch space in NekRS, which can then be applied for any purpose within the NekRS case files (both on the host and device). When modeling PTS, several factors can impact the results significantly. In this report, the impacts of the geometry of the model, Reynolds number and buoyancy effect are investigated. Two geometry, i.e., a simplified model and a realistic RPV model, with both laminar and turbulent flow condition are adopted for the PTS simulation with and without buoyancy effect. The purpose of the investigation is to understand the impact of these factors on the prediction of temperature history of RPV. The accurate prediction on the temperature evolution, which will be exported to Grizzly code for further analyses on the progression of aging mechanisms and their effects on the integrity of RPV structures, is very crucial. Based on the understanding of these factors, a more sophisticated model is built to analysis the PTS under SBLOCA scenario. A literature survey is conducted to pick the SBLOCA scenario for the multi-physics simulation. The analysis helps to explain the form and the transformation of the cold plum when the ECC is activated under SBLOCA. This model can be can be applied to study the PTS effect for different RPV configurations. The results can help to assess structural component degradation for advanced reactors.

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

Semi-elliptical underclad cracks resulting from the fabrication process of a reactor pressure vessel (RPV) were able to be detected by non-destructive testing method. Meanwhile, after long-term operation under severe conditions, such as high temperature, high pressure, and irradiation, the RPV becomes brittle and susceptible to damage, especially when subjected to pressurised thermal shocks (PTS). Therefore, the probabilistic fracture mechanics (PFM) analysis of RPV with the crack should be applied to evaluate the operation safety. To the best of the authors' knowledge, few studies or computer codes have applied PFM analysis for such cracks. Therefore, this study conducts PFM analysis for cracks by modifying the calculation procedure of FAVOR 12.1 computer code. The results show that during the lifetime of a nuclear power plant, such cracks will not threaten the RPV's safety. Additionally, three methods were proposed to improve FAVOR 12.1's ability to perform PFM analysis for axial through-clad cracking.

This paper presents an improved model for postulating fabrication flaws in reactor pressure vessels (RPVs) and for the treatment of measured flaw data by probabilistic fracture mechanics (PFM) codes that are used for structural integrity evaluations. The model used to develop the current pressurized thermal shock (PTS) regulations conservatively postulated that all fabrication flaws were inner-surface breaking flaws. To reduce conservatisms and uncertainties in flaw-related inputs, the United States Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) that has resulted in data on fabrication flaws from non-destructive and destructive examinations of actual RPV material. Statistical distributions have been developed to characterize the number and sizes of flaws in the various material regions of a vessel. The regions include the main seam welds, repair welds, base metal of plates and forgings, and the cladding that is applied to the inner surface of the vessel. Flaws are also characterized as being located within the interior of these regions or along the weld fusion lines that join the regions. Flaws are taken that occur at random locations relative to the embrittled inner region of the vessel. The probabilistic fracture mechanics model associates each of the simulated flaw types with the fracture properties of the region being addressed.

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

The current procedures for calculating pressure-temperature (P/T) limits for normal reactor startup and shutdown are defined by deterministic fracture mechanics methodology in the ASME Code, Appendix G (in both Section XI and Section III). The recent pressurized thermal shock (PTS) re-evaluation effort used a very thorough probabilistic fracture mechanics (PFM) evaluation to develop a technical basis to increase the PTS screening criteria. The feasibility of applying this same PFM methodology to evaluate normal startup and shutdown operation for both pressurized water reactor (PWR) and boiling water reactor (BWR) pressure vessels is described in this paper. The approach taken in this study was to define a new risk-informed margin term to be applied to the stress intensity factor for membrane tension (KIm) in Appendix G. The margin on KIm was determined by finding the value that results in a vessel failure frequency equal to 10−6 failures/reactor-year when the reactor operates up to the pressure-temperature limits calculated using the risk-informed margin. This simple approach was selected because it results in a minimum change to the current ASME Code procedure, is easy to apply when revising plant operating P/T limits, and allows for an increase in the allowable P/T limits by as much as would be provided by any alternate procedure. The results from this initial study indicate the margin on KIm in Appendix G can be reduced from 2 to 1.5 for PWR vessels for shutdown and potentially reduced from 2 to as low as 1 for startup. The results indicate the margin on KIm in Appendix G can be reduced from 2 to approximately 1 for BWR vessels for shutdown and potentially startup. Additional analyses for PWR and BWR vessels will be needed to develop a comprehensive risk-informed basis for any revisions to the ASME Code, Appendix G.

Application of the French Codes to the Pressurized Thermal Shocks Assessment