Beltline Region Research Articles

Ex-vessel neutron dosimetry results in the extended beltline region

Embrittlement of the pressure vessel has been a long-standing concern for pressurized water reactors. Historically, the beltline region of the pressure vessel, or the region of the pressure vessel located opposite the active fuel, has been the primary focus of embrittlement concerns. Embrittlement evaluations typically include a calculation of the neutron flux incident on the pressure vessel, which is in part validated with measurements. In the context of life extension, the beltline region, which is technically defined as the portion of the pressure vessel experiencing fast neutron fluence (E > 1.0 MeV) equal to or greater than 1017 n/cm2, is growing axially. Present fluence calculations for many plants indicate that the beltline region at the end of the 60 years of operation may extend to the bottom of the pressure vessel nozzles. Calculations in the extended beltline are difficult to validate as there are few dosimetry measurements available. At one Westinghouse 4-loop plant, ex-vessel neutron dosimetry sensor sets were irradiated in the extended beltline region. The measurements in the extended beltline region compare well with the discrete ordinates (RAPTOR-M3G) transport calculations and indicate the net uncertainty may be on the order of 30% in the extended beltline region.

Post-irradiation annealing of high flux irradiated and surveillance material reactor pressure vessel weld metal

In this study, high flux irradiated and surveillance high Ni and Mn and low Cu welds identical to those of the belt-line region of Ringhals R4 were subjected to annealing at temperatures between 390 and 455 °C for 24–30 h, in order to study the dissolution of irradiation induced clusters and possible matrix defects using hardness testing and atom probe tomography. It was found that the cluster characteristics did not change during annealing at 390 °C, meaning that the size, number density and composition of the clusters, which mainly consist of Ni and Mn, did not change. Thus, the observed decrease in hardness during annealing of the high flux irradiated material is believed to be due to dissolution of matrix defects that were stable at the operating temperature. Cluster dissolution was observed after annealing at 410 °C in the high flux irradiated material, leaving around 10% of the original clusters. These clusters contained more Cu and less Ni and Mn than before annealing. The cluster dissolution at temperatures above 400 °C correlated with the decrease in hardness. The larger clusters of the surveillance material required a higher temperature or longer time to be dissolved compared to the clusters of the high flux material.

Study on Fracture Assessment for BWR Reactor Pressure Vessel Steel by GTN model

In the JSME fitness-for service code, fracture evaluation methodology based on linear elastic fracture mechanics is provided for reactor pressure vessel (RPV) components by considering brittle fracture due to neutron irradiation in the RPV beltline region. However, in order to make a rational fracture evaluation for the RPV component such as bottom head component of boiling water reactor (BWR), where irradiation embrittlement is not concerned, it is necessary to consider the applicability of the ductile fracture evaluation methodology consistent with material fracture behavior. In this study, the applicability of the damage mechanics analysis method using Gurson-Tvergaard-Needleman (GTN) model, which is a numerical analysis method for ductile failure, was investigated for RPV bottom head component. Mechanical properties tests and fracture tests with plate specimens were performed using RPV law alloy steel (LAS) material. GTN model parameters of RPV LAS material that can reproduce stress-strain curve and load-displacement curve were determined. Fracture analyses of the plate specimens were performed using determined parameters as well. The load-Crack Mouth Opening Displacement (CMOD) curves and fracture behavior of the fracture tests were compared with the analyses results. The analyses predicted the maximum loads of the tests within an error of 4% and the fracture behavior between the tests and analyses was consistent. The results in this study clearly supported the applicability of the GTN model to RPV bottom head ring component.

Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock

This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.

Application of Flaw Updating Process on Probabilistic Integrity Analysis for a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks

The structural integrity of a reactor pressure vessel (RPV) is a crucial issue for an operating nuclear power plant, especially in the beltline region, which suffers the highest neutron irradiation. Owing to its capability of considering parameters based on statistical distributions and provision of objective risk-informed results, the probabilistic fracture mechanics (PFM) method is widely used in evaluating the structural integrity of RPVs. However, the flaw characteristics used for PFM analysis are mainly derived from particular vessel inspection information such as from the Pressure Vessel Research User Facility and Shoreham vessels, which may not be able to truly represent the vessel-specific condition of an analyzed RPV. In this work, the Bayesian inference, which combines prior flaw data with new inspection results as well as uncertainties, is used to develop posterior vessel-specific flaw distributions. Then, the updated flaw model is used for PFM analysis to investigate the effects on the fracture probability assessment of RPVs subjected to pressurized thermal shocks (PTSs). Considering the updated flaws based on the inspection data, the PFM analysis result could be more realistic to predict the fracture risks of RPVs during operation.

Development of pressure-temperature operation limits for a PWR vessel considering beltline shell and extended beltline nozzles

The pressure-temperature limits (P-T limits) of embrittled reactor pressure vessels (RPVs) should be determined for normal heat-up and cool-down operations based on fracture mechanics to ensure the structural integrity. In general, the P-T limits are mainly determined by the fracture toughness of beltline region material which suffers the highest level of neutron embrittlement. However, some other components such as nozzles which suffer insignificant neutron embrittlement may restrict the P-T limits as a result of higher stress concentration from their structural discontinuities. Therefore, beltline region material with the highest reference temperature as well as nozzles that have structural discontinuities need to be considered when developing the P-T limits of RPV. In this paper, the P-T limit calculations for a Taiwan domestic pressurized water reactor (PWR) vessel were performed considering both beltline shell and extended beltline nozzles according to the nonmandatory Appendix G to ASME Code Section XI. The three-dimensional finite element analyses were conducted to obtain the pressure and thermal stress distributions of extended beltline nozzles for P-T limits calculation. It is found that the beltline region still dominates the cool-down P-T limit, but the extended beltline region will partially control the heat-up limit. Further, the reference temperature ranges of extended beltline nozzles that may influence the P-T limits of RPV were also investigated. Present study is helpful for safe operation and also provides a reference for regulation of PWR plants in Taiwan.

Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

The American Society of Mechanical Engineers (ASME) Section III Rules for Construction of Nuclear Facility Components subsection NB-2331 Material for Vessels requires that the effects of irradiation shall be considered on material toughness properties in the core belt line region of the reactor vessel. The code also states that “the design specifications shall include additional requirements, as necessary, to ensure adequate fracture toughness for the service lifetime of the vessel.” In a design report of nuclear pressure vessel, the design and service loads do not include loads that are affected by fracture toughness of the material. However, in the cases of fitness-for-service assessment for component flaws (prevalent with age of component), irradiated material properties become highly relevant. An example of a fitness-for-service is that of a beyond design basis reactor vessel head drop accident in a pressurized water reactor with a nozzle junction flaw. As a case study, the critical size of a postulated external surface semielliptical circumferential crack in the combustion engineering three-loop pressurized water reactor nozzle–vessel junction is calculated using ansys Workbench (Academic version) with the applied impact load from the vessel head drop accident. Failure assessment diagrams for numerous crack depths and lengths were developed considering the fracture toughness properties of the irradiated reactor vessel steel. The mode I stress intensity results used in the failure assessment diagram were compared with the available finite element and the American Petroleum Institute (API) standard API 579 analytical solutions for validation, showing good agreement. From this case study, it is demonstrated that the effects of irradiation on fracture toughness become prominent at the same postulated crack size in the nozzle–vessel junction dispositioned as “safe” becomes “unsafe” in the fracture assessment diagram. Using the unified curve method, the irradiated fracture toughness data in design specification can be supplied so that it may be used in fitness-for-service analysis to account for component aging.

Numerical analysis of the BWR-5 beltline region containing a crack postulated for an extended period of operation

In the case of a BWR-5, the wall of the reactor vessel located adjacent to the core is named beltline and it is subjected to neutron embrittlement, lowering the fracture toughness and increasing the yield strength of the material. Cracks can be developed in the internal surface of this area. In this paper, a semi elliptical part through thickness crack has been postulated along the axial direction. Its depth is 1.25 inch which is 0.25 of the thickness of the wall vessel and its length is 6 times the crack depth (7.5 inches). The material of the vessel is a low alloy carbon steel.In order to consider an extended period of operation, the analysis must cover sixty years of operation. The normal conditions of operation and start-up and shutdown transients have to be taken into account. For this purpose, a numerical evaluation was done with the Finite Element Method in conjunction with linear elastic fracture mechanics. This analysis was compared with the results of the procedure under the scope of the Appendix G of the ASME Code Section XI and the NUREG-0800. The results were used to establish the pressure and temperature limits.

Probabilistic analysis of PWR Reactor Pressure Vessel under Pressurized Thermal Shock

The beltline region is the most important part of the reactor pressure vessel, become embrittlement due to neutron irradiation at high temperature after long-term operation. Pressurized thermal shock is one of the potential threats to the integrity of beltline region also the reactor pressure vessel structural integrity. Hence, to maintain the integrity of RPV, this paper describes the benchmark study for deterministic and probabilistic fracture mechanics analyzing the beltline region under PTS by using FAVOR code developed by Oak Ridge National Laboratory. The Monte Carlo method was employed in FAVOR code to calculate the conditional probability of crack initiation. Three problems from Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel (PROSIR) round-robin analysis were selected to analyze, the present results showed a good agreement with the Korean participants’ results on the conditional probability of crack initiation.

Phase and structural transformations in VVER-440 RPV base metal after long-term operation and recovery annealing

This study was carried out to evaluate the possibility of 1st generation VVER-440 reactors lifetime extension by recovery re-annealing with the respect to base metal (BM). Comprehensive studies of the structure and properties of BM templates (samples cut from the inner surface of the shells in beltline region) of operating VVER-440 reactor (after primary standard recovery annealing 475 °C/150 h and subsequent long-term re-irradiation within reactor pressure vessel (RPV)) were conducted. These templates were also subjected to laboratory re-annealing 475 °C/150 h.TEM, SEM and APT studies of BM after laboratory re-annealing revealed significant recovery of radiation-induced hardening elements (Cu-rich precipitates and dislocation loops). Simultaneously a process of strong phosphorus accumulation at grain boundaries occurs since annealing temperature corresponds to the maximum reversible temper brittleness development. The latter is not observed for VVER-440 weld metal (WM).Comparative assessment of the properties return level for the beltline BM templates after recovery re-annealing 475 °C/150 h showed that it does not reach the one typical for beltline WM after the same annealing.

Pressurized Thermal Shock analysis of the reactor pressure vessel

PTS analysis is a substantial part of the technical analyses supporting the license application for operation beyond design life. Guidance for the analysis’ concept is given by the Hungarian nuclear regulator which is in line with the relevant international rules. Involving technical support organizations and independent consultants as well as taking into account the changes in fuel management Paks NPP, Hungary, compiled a new PTS analysis. Subjects of the analysis were the RPV belt line region components as well as other circumferential welds of the RPV including nozzle region. Beyond the PTS initiating events selected on the basis of engineering judgment a PSA provided additional transients which showed a higher frequency than 10-5 /year. Thermal-hydraulic calculations completed for each selected PTS initiating event provided the necessary input for the structural analysis. Based on core configurations end-of-life fluence were calculated. Neutron dosimetry surveillance results were used to verify the calculations. Temperature and stress field calculations were performed by solving the system of equations of elasticity. Underclad crack was postulated for fracture mechanics calculation. KI was calculated at the crack tip and the boundary between the cladding and base / weld metal; KIc was calculated on the basis of the reference curve. Comparison of these two parameters (KI ≤KIc), i.e. evaluation of whether the postulated defect stability criteria was met, gave the result.

Ultimate Bearing Capacity Analysis and Sizing Optimization of a Cracked Reactor Pressure Vessel in the Coupled Thermo-Mechanical Field

Pressurized thermal shock (PTS) can subject a crack surface to a very high tensile stress. Also the material toughness is obviously decreased in the cooling process, so it is necessary to study the influence of PTS on the ultimate bearing capacity of a reactor pressure vessel with defects. A 3-D finite element model is established for the beltline region around an inner crack. The FEM is used to reveal the transient temperature field and stress field, and the XFEM is adopted to simulate the ductile crack propagation. To ensure that the strength requirement is satisfied, the ultimate internal pressures of vessels with different crack sizes and different wall thicknesses are obtained. The result shows that the ultimate bearing capacity of the base wall with shallow surface cracks at high temperature is mainly controlled by tensile strength, while it is also affected by the fracture toughness of the material under the severe PTS. The stress in the early stage of the PTS is mainly the thermal stress, and later is the thermo-mechanical coupling stress. The impact of the crack depth on the bearing capacity of the structure is much greater than that of the crack length.

Elastic and elastoplastic fracture analysis of a reactor pressure vessel under pressurized thermal shock loading

Abstract Due to the limitation of the manufacturing process and improper operation, there may be defects near the inlet nozzles of reactor pressure vessel (RPV). Under pressurized thermal shock (PTS), the crack region is subjected to a high tensile stress, and the material toughness is gradually decreased in the cooling process. So it is necessary to evaluate the integrity of RPV in the thermo-mechanical coupling field. The 3-D finite element model is established for the beltline region around the inlet nozzles of a RPV. According to the calculated nil-ductility reference temperature, the elastic constitutive relation is firstly established for the fracture analysis. The stress intensity factors along the crack tips are compared with the material toughness. Using the XFEM, the crack propagation path is obtained to verify the predicted results. At relatively low reference temperatures, the ductility characteristics of the nuclear material can be exhibited. In this case, the ultimate bearing capacity of the structure is demonstrated by the elastoplastic fracture analysis. Finally, the safe range of reference temperature is summed up to prevent the surface crack from running through the whole wall thickness of the vessel during the PTS transient.

Ultimate bearing capacity analysis of a reactor pressure vessel subjected to pressurized thermal shock with XFEM

Abstract Under pressurized thermal shock (PTS), the structure integrity of reactor pressure vessel (RPV) meets a potential challenge. The vicinity of inlet nozzle is the weak link and hazardous area of the vessel. Once crack initiation occurs here, the stress concentration around the crack tip would lead to plastic deformation, crack propagation, and even the failure of the whole structure. Based on ABAQUS, the beltline region around the inlet nozzle was modeled to calculate the response of the transient temperature field and stress strain field. In order to demonstrate the ultimate bearing capacity of the RPV with the critical crack sizes, the dynamic process of crack propagation was simulated with the XFEM. Through the quantitative analysis of the Mises stress and circumferential stress through the wall thickness, the crack propagation paths in different directions were confirmed. The above analysis and calculation provide an important reference for the safety and reliability assessment of RPV.

Monte Carlo evaluation of neutron irradiation damage to the VVER-1000 RPV

Neutron irradiation damage is of the most critical damage mechanisms in nuclear power plant's pressure vessel. Neutrons transfer their kinetic energy to target atoms of RPV that start to jump creating vacancies and interstitials known as Frenkel Pair (FP). The FPs are responsible for formation of defected clusters and microstructural modifications (e.g. phase reactions, segregations). These effects deteriorate physical and mechanical properties of the RPV steels among which are increasing the hardness and decreasing the embrittlement which results in limiting the life-time of RPV. The most sensitive location in the RPV related neutron irradiation damages is in the beltline region and adjacent to the reactor core. Welds and their heat affected zones (HAZs) in this region are of particularly importance due to higher probability flaws compared with the base metal. In this paper, Monte Carlo simulation of the detailed reactor core have led to identifying the areas maximum neutron flux in the RPV. Then the SPECTER and SRIM Monte Carlo codes are used to evaluate the neutron spectral-averaged DPA values for the beltline region of RPV.

Evaluation of radiation damage in belt-line region of VVER-1000 nuclear reactor pressure vessel

Regarding irradiation damage, the most sensitive region in the reactor pressure vessel (RPV) is the area adjacent to the reactor core; the so called belt-line region. Belt-lines and their related heat affected zones (HAZs) in the belt-line region are constantly under neutron and gamma irradiation and have a higher potential to various flaws. The microscopic defects produced in materials due to irradiation are referred to as radiation damage; the units of displacements per atom (DPA) are used routinely to characterize the extent of such damage. In this research, for a typical VVER-1000 (V446) RPV, the areas of the highest neutron radiation flux (belt-line) is first identified by using a Monte Carlo neutron transport code. This is used to calculate and validate the neutron flux and fluence at the belt-line. Employing the PTRAC output from MCNP and the SRIM Monte Carlo code, radiation damage in the RPV is simulated and assessed in terms of DPA. The integrated damage for a full effective year is evaluated to be order of 5E-04 displacements per cm thickness of RPV belt-line metal.

Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 in. below to approximately 12 in. above the height of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

The radiation embrittlement behavior of reactor pressure vessel shell is influenced by the chemistry concentration of metal materials. This paper aims to study the effects of copper and nickel content variations on the fracture risk of pressurized water reactor (PWR) pressure vessel subjected to pressurized thermal shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel assumed with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the U.S. NRC's new PTS rule are applied as the loading condition. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.