in use: bwr and pwr irradiated materials testing and...

Materials Degradation and Aging August 2015

IN USE: BWR AND PWR IRRADIATED MATERIALS TESTING AND DEGRADATION MODELS FOR REACTOR INTERNALS

ISSUE STATEMENT

The performance of BWR and PWR reactor internals is affected by several irradiation-based degradation mecha-nisms: irradiation–assisted stress corrosion cracking (IASCC), irradiation embrittlement (reduction in ductility and fracture toughness), creep, stress relaxation and void swelling. Limited understanding of the factors affecting these degradation modes (such as dose, dose rate, stress intensity factors, effects of specimen size and orientation, solute additions, temperature and environmental conditions) could impact decisions related to extended operation up to 80 years. Knowledge gaps exist, for example, on the effects of high fluence on degradation of base metal, welds and heat affected zones in BWR and PWR environments. Experi-mental information is also lacking that could lead to the development of more robust, fundamentally based models and radiation resistant materials.

DRIVERS

Asset ManagementImproved understanding of irradiation effects can signifi-cantly affect decisions related to repairs, replacements and even overall unit life. Improved understanding of irradia-tion-induced degradation based on additional data and more accurate models can be used to make sound repair or replace-ment decisions and avoid unanticipated outages associated with cracking of internals.

Regulatory and Industry CommitmentsRegulatory commitments during license renewal typically require plants to implement an aging management program that follows industry guidance. Such programs are based on the technical foundation established in the inspection and evaluation guidelines developed and maintained through EPRI research. This work also supports industry commit-ments through efforts such as the NEI-03-08 Materials Initiative.

RESULTS IMPLEMENTAT ION

The improved data and degradation models developed under this program will be incorporated into industry guidance for managing internals degradation in BWRs and PWRs. The goal is to provide all necessary data and models to support light water reactor operation through 80 years. Examples include:• Improved crack growth rate models and disposition curves

for BWR and PWR internals based on an expanded data-base, input from an expert panel and ASME code/regula-tory approval. These models will support utility decisions on inspection frequency, repair and replacements

• Improved IASCC initiation models for PWR internals will be developed based on data from crack initiation tests on materials removed from the retired PWR plants such as Zorita and in-reactor crack initiation tests with peri-odic dynamic loading in the Halden reactor. MRP will also perform crack initiation tests in autoclaves to study the effect of lithium and investigate the effect of dynamic loading on crack initiation (with EDF). The materials include stainless steel base metals, welds and heat affected zones.

• Improved fracture toughness models for BWR and PWR internals to support structural margin and integrity analyses.

• Improved void swelling model for PWRs benchmarked with results from the Gondole project and further vali-dated through a comparison with other void swelling codes

• Radiation-resistant materials for future replacements of internals in existing plants or for new plants (Advanced Radiation Resistant Materials Program)

• Development of fundamentals models to link microstruc-ture of irradiated materials with their engineering proper-ties e.g., yield strength, ductility, fracture toughness and IASCC susceptibility

PROJECT PLAN

Long-term irradiation effects will be characterized by crack initiation and crack growth tests on materials removed from retired plants such as the Zorita PWR in Spain. The Zorita plant was operated for 38 years (26 Effective Full Power Years) and the highest accumulated fluence on the reactor vessel internals is approximately 50 displacements per atom

EPRI | Nuclear Sector Roadmaps August 2015

(dpa). Some of the material removed from Zorita will be re-irradiated in a research reactor to achieve fluence values of approximately 65 dpa expected in 60 to 80 years and tested.

EPRI will continue its participation in the international col-laborative program on crack initiation and crack growth studies of irradiated materials at the Halden Reactor. EPRI will also continue the collaboration with the Idaho National Laboratory to conduct irradiations and post-irradiation test-ing of alloys X-750 and XM-19 at the Advanced Test Reactor.

An EPRI expert panel has developed an improved IASCC crack growth model for BWR and PWR internals after com-piling and screening an expanded database. The model can be used predict residual lifetimes of irradiated internals through license extension. The models will be submitted for code and regulatory review.

EPRI has initiated new TI break-through project on Rapid Simulation of Irradiation Damage in stainless steels The goal of this project is to develop and validate a cost effective method to simulate irradiation induced degradation (e.g., void swelling) at high fluence. EPRI and DOE are collabo-rating on modeling the relationship between localized defor-mation and IASCC response.

RISKS

If this work is not done, conservative crack growth and frac-ture toughness curves for flaw evaluations may be applied that result in more frequent inspections and unnecessary radiation exposure and costs.

Achieving progress will depend on the availability of irradi-ated test materials and data that are representative of inter-nals in service and have accumulated fluence levels expected during license extension. Obtaining such materials is a sig-nificant challenge because truly appropriate materials are rarely removed from operating reactors, relatively few plants are permanently shut down, and acquisition is often chal-lenging and expensive.

The accessibility of some components may limit the effec-tiveness of current examination methods to provide reliable inspections and capture information related to damage mechanisms. This type of information is needed to validate the degradation models.

The internal materials currently in service are susceptible to irradiation-induced degradation and there is a need to develop more radiation-resistant materials if replacements become necessary. Otherwise, utilities will continue to incur increased costs of inspections and repairs of internals in the future.

RECORD OF REVISION

This record of revision will provide a high level summary of the major changes in the document and identify the Road-map Owner.

revision description of change

0 Original Issue: August 2011 Roadmap Owner: Raj Pathania

1 Revision Issued: August 2012 Roadmap Owner: Raj Pathania

Changes: Added Key Milestones to the flowchart


Changes: Added additional tasks and completed and future milestones to the flow chart

3 Revision Issued: November 2013 Roadmap Owner: Raj Pathania

Changes: Revised schedule and funding status of MRP task on IASCC Program on Baffle Bolts


Changes: Updated text. Revised schedule of BWRVIP tasks on BWRVIP-100 revision and INL tests. Updated schedule of MRP Zorita tasks. Added PSCR task on code and regulatory review of IASCC models and revised schedule of task on size effect study with Zorita material


Changes: Updated text. Updated status and schedule of existing tasks. Added new tasks on rapid simulation of irradiation damage, additional studies on effect of lithium on IASCC and thermal aging analysis of stainless steel welds.

Materials Degradation and Aging August 2015

EPRI | Nuclear Sector Roadmaps August 2015

in use: bwr and pwr irradiated materials testing and...

Documents