Beltline Materials Research Articles

Evaluation of RPV closure head penetration integrity considering fitting and residual stresses

To ensure the structure integrity of the Reactor Pressure Vessel (RPV), the main challenge is the embrittlement of beltline material. However, the stress is more complex for the regions of Closure Head Penetration (CHP). Feedback of leakage at the CHP has been found during in-service inspection. Although the detailed mechanical evaluation of the RPV head and the CHP has been carried out by many international institutions, there are limited studies to concern structure integrity of CHP tube. The systematic analysis of the influence of material strength matching, assemble stress, the Welding Residual (WR) stress and work pressure load on the fracture is also limited. In this paper, the systematic influence of these factors is analyzed, and the distribution characteristics of Stress Intensity Factor (SIF) and the crack propagation characteristics are studied. The 2D model and accurate 3D model are carried out using the finite element method. The results show that the WR stress on the outer surface of the CHP and the area adjacent to the weld is larger than the yield strength of the material. There is a gap between the RPV and the upper part of the CHP. The attenuation rate of the WR stress in the outlet direction of RPV is lower than that in the inlet direction, so cracks are more likely to occur in the out of direction. The influence of WR stress on the circumferential stress distribution and fracture parameters in the wall thickness direction of CHP is bigger than the influence of assemble load and design condition load. However, the overall stress of CHP is low. The SIF values along the crack front have an asymmetric distribution, and the SIF value is very small. As the fatigue crack propagation life is much longer than the assumed limit in the design lifetime, the risk of fracture and crack fatigue propagation caused by mechanical stress is low.

Fracture Assessment for Pressurized Water Reactor Vessel Nozzles Subjected to Pressure and Thermal Loading

Abstract To ensure the structure integrity of the reactor pressure vessel (RPV) in pressure water reactor plants, the main challenge is the embrittlement of beltline materials. However, the stress at inlet or outlet nozzles of the RPV which are reinforced in comparison with the beltline is more complex, especially under the thermal loads. In recent studies, significant research efforts have been conducted to show that the nozzle region may be more challenging under some conditions. In this paper, a fracture assessment for the RPV nozzles subjected to pressure and thermal loading is studied using the software abaqus 6.12 and Zen Crack 7.9-3. It includes: stress intensity factor calculation based on three-dimensional finite element method; structural integrity assessment under a typical loss of coolant accident (LOCA) transient; and the fatigue crack growth evaluation under cyclic loading situations. The results show that the stress intensity factor along the crack front is evidently asymmetric, and only to assess the safety of the deepest point along the crack front in the ASME and RCC-MR codes may be unreasonable. If the KIa criteria is applied, under a typical LOCA transient, it is difficult to obtain an effective fracture safety margin for a 1/4 thickness crack, while based on the KIC criteria, the nozzle is shown to be safe in the case study. The long shallow surface crack (which is often easily produced in the nozzle area) tends to be circle type under the cyclic pressure loading situation which shows that the crack shape assumed in the ASME and RCC-MR codes is reasonable.

The probabilistic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

The pressurized thermal shock (PTS) event poses a potentially significant challenge to the structural integrity of the reactor pressure vessel (RPV) during the long time operation (LTO). In the USA, the “screening criteria” for maximum allowable embrittlement of RPV material, which forms part of the USA regulations, is based on the probabilistic fracture mechanics (PFM). The FAVOR software developed by Oak Ridge National Laboratory (ORNL) is used to establish the regulation. As the technical basis of FAVOR is not the most widely-used and codified methodologies, such as the ASME and RCC-M codes, in most countries (with exception of the USA), proving RPV integrity under the PTS load is still based on the deterministic fracture mechanics (DFM). As the maximum nil-ductility-transition temperature (RTNDT) of the beltline material for the 54 French RPVs after 40 years operation is higher than the critical values in the IAEA-TECDOC-1627 and European NEA/CSNI/R(99)3 reports (while still obviously lower than the “screening criteria” of the USA), it may conclude that the RPV will not be able to run in the LTO based on the DFM. In the FAVOR, the newest developments of fracture mechanics are applied, such as the warm pre-stress (WPS) effect, more accurate estimation of the flaw information and less conservation of the toughness (such as the three-parameter Weibull distribution of the fracture toughness). In this paper, the FAVOR software is first applied to show both the methodology and the results of the PFM, and then the limits in the current FAVOR software (Version 6.1, which represents the baseline for re-assessing the regulation of 10 CFR 50.61), lack of the impact of the constraint effect, ductile–tearing initiation failure analysis, elastic–plastic fracture mechanics analysis, safety assessment of the crack front interface points and interaction among multiple cracks, are also discussed.

Analysis of the Belgian Surveillance Fracture Toughness Database Using Conventional and Advanced Master Curve Approaches

Abstract The “classical” regulatory approach to the analysis of surveillance capsules in nuclear power plants entails an indirect estimate of the fracture toughness of the beltline materials, by inferring rather than measuring their toughness properties. Indeed, the irradiation-induced shift of the fracture toughness curve is assumed to be equal to the shift of the Charpy absorbed energy transition curve at a predefined level (41 J). An alternative surveillance approach, primarily based on direct fracture toughness measurements in the ductile-to-brittle transition region using the Master Curve procedure, has been applied to surveillance materials from several Belgian nuclear power plants in the past 15 years. This has led to the establishment of a significant database, consisting of 292 fracture toughness data points for 23 material conditions (unirradiated materials and surveillance capsules). In this study, different temperature normalization approaches are applied to the available data. The analyses show that data clearly follow the Master Curve formalism. Moreover, it is confirmed that both the static (KIc) and the dynamic (KIR) curves of the ASME Code Section XI provide an effective lower bound to the measured results, although more conservatism is evident when using RTNDT as the normalization parameter. Both the conventional (ASTM E1921-08, ‘‘Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range’’) and advanced (Multi-Modal) Master Curve analyses of the database clearly demonstrate that normalizing data by (T−RTT0) provides the best rationalization of the available information and the most effective representation of the experimental scatter.

Using Subsize Impact Bend Specimens for Estimation of Irradiation and Re-Irradiation Embrittlement of VVER RPV Steels

Abstract Results of impact bend tests of subsize Charpy V-notch-type (CVN) specimens are used in Russia for monitoring both re-irradiation embrittlement of the first generation of VVER-440 (including VVER-440/179 and VVER-440/230, i.e., projects B-179 and B-230) reactor pressure vessel (RPV) steels and irradiation embrittlement of VVER-440/213 RPV steels. For the first generation of VVER-440 RPVs, no surveillance program was foreseen and only subsize specimens (5 by 5 by 27.5 mm and 3 by 4 by 27 mm) can be manufactured from samples taken from the inner surface of the RPV for radiation embrittlement monitoring. For VVER-440/213 RPVs, 80 % of the surveillance specimens are irradiated with a lead factor greater than 20. Only four specimens of base metal and four specimen halves of weld seam material are irradiated with a lead factor less than 3. Taking into account that full-size (10 by 10 by 55 mm) CVN specimens were used in surveillance programs, using subsize specimens enables to solve the flux effect problem for VVER-440/213 RPV beltline materials and ensure representativity of surveillance programs. Application of subsize specimens for the estimation of radiation embrittlement requires developing a correlation between transition temperature values evaluated using test results of standard (full-size) and subsize CVN specimens. Such a correlation is discussed.

Application of Master Curve fracture toughness for reactor pressure vessel integrity assessment in the USA

The Master Curve fracture toughness approach has been used in the USA for better defining the transition temperature fracture toughness of irradiated reactor pressure vessel (RPV) steels for end-of-life (EOL) and EOL extension (EOLE) time periods. The first application was for the Kewaunee plant in which the life-limiting material was a circumferential weld metal. Fracture toughness testing of this weld metal corresponding to EOL and beyond EOLE was used to reassess the PTS screening value, RT PTS, and to develop new operating pressure–temperature curves. The NRC has approved this application using a shift-based methodology and higher safety margins than those proposed by the utility and its contractors. Beaver Valley Unit 1, a First Energy nuclear plant, has performed similar fracture toughness testing, but none of the testing has been conducted at EOL or EOLE at this time. Therefore, extrapolation of the life-limiting plate data to higher fluences is necessary, and the projections will be checked in the next decade by Master Curve fracture toughness testing of all of the Beaver Valley Unit 1 beltline materials (three plates and three welds) at fluences near or greater than EOLE. A supplemental surveillance capsule has been installed in the sister plant, Beaver Valley Unit 2, which has the capability of achieving a higher lead factor while operating under essentially the same environment. The Beaver Valley Unit 1 evaluation has been submitted to the NRC. This paper reviews the shift-based approach taken for the Beaver Valley Unit 1 RPV and presents the use of the RT T 0 methodology (which evolved out of the Master Curve testing and endorsed through two ASME Code Cases). The applied margin accounts for uncertainties in the various material parameters. Discussion of a direct measurement of RT T 0 approach, as originally submitted for the Kewaunee case, is also presented.

Annealing of nuclear reactor pressure vessels

The neutron embrittlement that occurs in the beltline of reactor pressure vessels (RPV) can be managed by various techniques such as fuel management, but only thermal annealing can reverse the effects and result in a restoration of RPV beltline material toughness. The US Nuclear Regulatory Commission has recently revised the Code of Federal Regulations to include the use of thermal annealing of RPV for recovery of material toughness. The Annealing Rule, 10 CFR Part 50.66, has an associated Regulatory Guide 1.162 that describes the format and content of a thermal anneal report that must be submitted to the NRC prior to performing an anneal. This paper will describe the thermal annealing process including regulatory requirements in 10 CFR Part 50.66, techniques for predicting and measuring the toughness recovery, and NDE requirements. Although 14 Russian-designed RPVs have been annealed, there are sufficient differences between the Russian and US designs to question the ease of thermal annealing without producing any unwanted dimensional changes in the RPV and associated piping. The paper will discuss the ongoing annealing demonstration project supported by the Department of Energy which performed a thermal anneal on a canceled pressured water reactor at Marble Hill, Indiana. The associated NRC programs also will be described. This annealing demonstration will be used to bench mark the expected thermal and stress distributions created by thermal annealing and minimize the possible dimensional changes in the RPVs. The paper also will discuss the first possible implementation of thermal annealing for a US commercial nuclear power plant and some important issues that will need to be addressed.

Reactor pressure vessel integrity program

The reactor vessel report provides a technical evaluation of the effects of aging on the reactor vessel shell, lower vessel head, closure head, nozzles (and safe ends, if provided), interior attachments, and all associated pressure-retaining bolting. The evaluation is performed in accordance with the requirements of the license renewal rule, 10 CFR Part 54. This paper is based upon the B and W Owners Group report BAW-2251 (Mackay, W.H., Gross, L.B., Rinckel, M.A., Starkey, R.L., 1996. Demonstration of the management of aging effects for the reactor vessel. Report BAW-2251, Framatome Technologies, Lynchburg, VA), which was submitted to the NRC for review and approval in June 1996. The reactor vessel report demonstrates that existing utility programs adequately monitor and manage the relevant effects of aging of the reactor vessel such as cracking, loss of material, reduction in the fracture toughness of the vessel beltline material, and loss of mechanical closure integrity. The aging management programs credited include ASME Section XI, ISI, compliance to 10 CFR Parts 50.60 and 50.61, Plant Technical Specifications, and commitments to generic Nuclear Regulatory Commission communications. The aging management techniques described in this report permit continued safe operation over the period of extended operation associated with license renewal. In addition, the report includes the evaluation of three generic time-limited aging analyses: (1) fatigue of metallic components; (2) analyses and calculations performed to show compliance with NRC regulations governing reduction of fracture toughness of reactor vessel beltline materials; and (3) intergranular separations in low-alloy steel heat-affected zones under austenitic stainless steel cladding.

Use of J– R curves in assessing the fracture behavior of low upper shelf toughness materials

The objective of this investigation was to evaluate the use of small specimen J– R curves in assessing the fracture resistance behavior of reactor vessels containing low upper shelf (LUS) toughness weldments. As required by the U.S. Code of Federal Regulations (10 CFR, Part 50), reactor vessel beltline materials must maintain an upper shelf Charpy V-Notch (CVN) energy of at least 50 ft-lbs (68 J) throughout vessel life. If CVN values from surveillance specimens fall below this value, the utility must demonstrate to the U.S. Nuclear Regulatory Commission (NRC) that the lower values will provide “margins of safety against fracture equivalent to those required by Appendix G of the ASME Boiler and Pressure Vessel Code”. This paper will present recommendations regarding the material fracture resistance aspects of this problem and outline an analysis procedure for demonstrating adequate fracture safety based on CVN values. It is recommended that the deformation formulation of the J-integral be used in the analysis described above. For cases where J-integral fracture toughness testing will be required, the ASTM E1152-87 procedure should be followed, however, data should be taken to 50% to 60% of the specimen remaining ligament. Extension of the crack growth validity limits for J– R curve testing, as described in E1152-87, can be justified on the basis of a “ J-controlled crack growth zone” analysis which shows an engineering basis for J-control to 25% to 40% of the specimen remaining ligament. If J- R curve extrapolations are required for the analysis, a simple power law fit to data in the extended validity region should be used. The example analysis performed for low upper shelf weld material, showed required CVN values for a reactor vessel with a 7.8 inch (198 mm) thick wall ranging from 32 ft-lbs (43 J) to 48 ft-lbs (65 J), depending on the magnitude of the thermal stress component.

A review of the ASME low upper shelf toughness evaluation procedures for nuclear reactor pressure vessels

The ASME Section X1 Working Group on Flaw Evaluation has proposed criteria for the evaluation of reactor pressure vessel beltline materials which have an upper shelf energy less than 50 ft-lbs (69 J). These criteria have been assessed and applied to Linde 80 weld materials in recent investigations; this assessment and evaluation are described in the paper. A key element in the evaluation procedure is the J R curve for the relevant material. Recent experimental studies have demonstrated that the J R curve is size dependent for some materials, in the sense that the J R curve slope decreases with increasing specimen thickness. This paper assesses this experimental work and discusses it in the context of the integrity of nuclear reactor pressure vessels.