Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Univ. of Idaho, Moscow, ID (United States)
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically decrease run times.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1244643
- Report Number(s):
- INL/EXT-15-37121; M3LW-15IN07040615; TRN: US1601167
- Country of Publication:
- United States
- Language:
- English
Similar Records
SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS
Modular system for probabilistic fracture mechanics analysis of embrittled reactor pressure vessels in the Grizzly code
Related Subjects
97 MATHEMATICS AND COMPUTING
PRESSURE VESSELS
PROBABILISTIC ESTIMATION
FRACTURES
DEFECTS
NUCLEAR POWER PLANTS
PROBABILITY DENSITY FUNCTIONS
FRACTURE PROPERTIES
EVALUATION
TRANSIENTS
STRESS INTENSITY FACTORS
BOUNDARY CONDITIONS
CRACK PROPAGATION
RISK ASSESSMENT
THERMAL SHOCK
BENCHMARKS
INTEGRALS
SAMPLING
COMPUTERIZED SIMULATION
R CODES
G CODES
EMBRITTLEMENT
Fracture Mechanics
Pressurized Thermal Shock
Probabilistic
Reactor Pressure Vessel