May-June 2019 NPJ

46 NuclearPlantJournal.com Nuclear Plant Journal, May-June 2019 3. How do you plan to extend the life of existing reactors through improved prediction of the lifetimes of key structural components? CASL is pursuing linkage of VERA with the age-related degradation code Grizzly, which was developed under the DOE LWR sustainability program to address the breadth of material aging issues for nuclear reactors. In order to model the aging process, VERA provides as input to Grizzly, the highly resolved dose and reactor operating temperature history for any reactor component which is critical to material embrittlement and age-related degradation. This includes not only the reactor pressure vessel where the concern is pressurized thermal shock but potentially any structural component, such as concrete, where long-term neutron or gamma irradiation is a concern. The goal is an advanced modeling and simulation capability that can be utilized for analyses in support of Subsequent License Renewal that would extend operation of the existing reactor fleet beyond a 60-year lifetime. 4. How will CASL ensure safety of the plant to prevent an accident similar to Fukushima, TMI and Chernobyl? CASL has several challenge problems related to enhancing the safety and understanding of nuclear reactors during accident scenarios that include the loss of coolant accident (LOCA), the reactivity insertion accident (RIA) and departure from nucleate boiling (DNB). In all cases, the concern is maintaining the integrity of the fuel form during the accident progression. For example, the LOCA capability includes advanced models for cladding burst failure, axial fuel relocation, transient fission gas release, and energy deposition arising from rapid clad oxidation. RIA is focused on the accurate modeling of the reactor core response and total energy deposition in the fuel through the progression of the event that includes a rapid power ascension and its turnaround through negative fuel temperature feedback. The DNB capability is based on advancements in two-phase flow modeling within the framework of computational fluid dynamics that has enabled virtual simulations for the condition of clad surface dry out (critical heat flux). VERA capabilities have contributed to advancing the understanding of postulated accident progressions, an example being the main steam line break (MSLB) analysis displayed in Figure 2. Shown is the VERA simulation of core power distribution for the MSLB condition of hot zero power with offsite power available, which was found to be the limiting scenario with respect to DNB acceptance criterion consistent with the limiting analysis of the Westinghouse 4-loop plant safety analysis report. These advanced predictive capabilities within VERA have been applied to existing fuel as well as accident tolerant fuel (ATF). The use of VERA will assist in bringing ATF to market more quickly while adding another layer of protection to the existing conservative industry designs. 5. How was your project implemented at Watts Bar 1 and 2 and what results were accomplished. CASL has been validated for all fuel cycles of the Watts Bar 1 nuclear reactor (cycles 1-16) and shown to be in excellent agreement with measured data from the entire operating history. Watts Bar 1 has served as the test bed for VERA capability development as pertains to the CASL challenge problems as well as new applications being developed. Working with TVA, VERA has been used for a wide range of applications that includes core design, operational issues, outage support, and plant lifetime analysis. Examples of specific analyses include investigation of radial power distribution and flow anomalies, vessel fluence benchmark over the core lifetime, secondary source design to support plant outages, and crud analysis to assess the risk of axial offset anomaly during reactor operation. For Watts Bar 2, the first commercial power reactor startup in two decades, CASL provided blind predictions to both Westinghouse and TVA of the reactor core behavior in advance of the initial criticality, which Westinghouse used to identify areas of improvement within their commercial tools. Comparisons with the startup physics were excellent and confirmed the reliability and accuracy of the VERA in a production environment. VERA was used to follow the detailed power escalation and testing history in near real-time using the Oak Ridge Leadership Computing Facility. With Watts Bar 2 now in its third cycle of operation, VERA has continued to be used as a tool for analysis of core behavior. 6. Other. As a CASL founding partner, Westinghouse has been integral to the development of VERA and has used VERA in a number of different application areas. The most recent has been the startup analysis of the first AP1000 reactors at Sanmen which, similar to Watts Bar 2, was performed as a blind prediction. The AP1000 core designs were first-of-a-kind with highly heterogeneous fuel designed to maximize fuel cycle efficiency. VERA comparison of predicted results with global and local measured parameters showed excellent agreement and reinforced confidence in the startup predictions achieved using currently licensed industry methods. Westinghouse has applied and continues to assess the use of VERA across a broad range of application areas that include safety analysis (main steam line break and reactivity insertion accidents), reactor operations (margin assessment of reactor trip setpoints), core design (modeling improvements and crud analysis for CIPS), and fuel performance of ATF. Contact: Jason Ellis, Oak Ridge National Laboratory, telephone: (865) 241-4819, email: ellisjk@ornl.gov . CASL... ( Continued from page 31)