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LWR Sustainability (LWRS) Program Status Uprate. Dr. Hongbin Zhang Idaho National Laboratory Hongbin.Zhang@inl.gov 208-526-9511 . International RELAP5 Users Group Meeting and Seminar Salt Lake City, July 25-28, 2011. LWRS Vision, Goals, and Scope. Vision

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slide1

LWR Sustainability (LWRS) Program Status Uprate

Dr. Hongbin Zhang

Idaho National Laboratory

Hongbin.Zhang@inl.gov 208-526-9511

International RELAP5 Users Group Meeting and Seminar

Salt Lake City, July 25-28, 2011

lwrs vision goals and scope
LWRS Vision, Goals, and Scope

Vision

  • Enable existing nuclear power plants to safely provide clean and affordable electricity beyond current license periods (beyond 60 years)

Program Goals

  • Develop fundamental scientific basis to allow continued long-term operation of existing LWRs
  • Develop technical and operational improvements that contribute to long-term economic viability of existing nuclear power plants

Scope

  • Materials Aging and Degradation
  • Risk-Informed Safety Margin Characterization
  • Economics and Efficiency improvements
  • Advanced Instrumentation and Controls
  • Advanced Fuels Development
extending the service life of today s lwr fleet may create new material challenges
The LWRS R&D scope provides the scientific basis for understanding and predicting materials aging and degradation within components, systems, and structuresExtending the service life of today’s LWR fleet may create new material challenges
  • Reactor metals (RPV’s, internals, steam generators, balance of plant, and weldments)
    • Mechanisms of IASCC
    • High-fluence effects on RPV steel
    • Crack initiation in Nickel based alloys
  • Concrete
    • Concrete aging for long term operation
    • Monitoring tools for concrete
  • Buried piping
    • Assessment on long term piping performance
  • Cabling
    • Assessment of cable aging issues
  • Mitigation, repair, and replacement technologies
    • Weld repair techniques
    • Post irradiation annealing
    • Advanced replacement alloys
fy11 lwrs materials aging and degradation research encompasses the entire power plant
FY11 LWRS Materials Aging and Degradation research encompasses the entire power plant

Concrete degradation

Analysis of cable degradation

Repair welding

Thermal annealing

Crack initiation in Ni-base alloys

Mechanisms of IASCC

Swelling of core internals

High fluence effects on RPV

Surrogate materials and attenuation

High fluence phase transformations

Advanced replacement alloys

Buried piping analysis

slide6
In addition to other research for extended service, research must also identify other or new topics before they become life-limiting
  • “Knowing the unknowns” is a difficult problem that must be addressed.
  • This is a particularly difficult issue for such a complex and varied material/environment system.
  • An organized PMDA approach is being employed.
  • Together with the USNRC, the LWRS program is working to expand the initial PMDA activity (NUREG 6923) to encompass broader systems and
  • longer lifetimes
    • Core internals and primary piping
    • Pressure Vessel
    • Concrete
    • Cabling

Proactive Materials Degradation Assessment Matrix

doe epri ceng nuclear plant life extension demonstration npled project
DOE/EPRI/CENG Nuclear Plant Life Extension Demonstration (NPLED) Project
  • In August 2009, Industry LTO Advisors asked EPRI to identify a generic pilot plant to investigate license renewal issues
  • Constellation Energy (CENG) offered Ginna Station and Nine Mile Point Unit 1 Station as industry pilot plants
  • Activity will be collaboration among U.S. DOE, EPRI, and CENG, to be conducted through DOE LWRS and EPRI LTO Projects
  • Project has 3 year planned duration with option for extension
  • Scope:
    • Reactor Vessel Internals Enhanced Aging Inspection
    • Comprehensive Containment Assessment
    • Investigation of Surveillance Samples for Projecting Reactor Vessel Life
    • Medium and low voltage power cable aging in adverse environments
    • Buried piping degradation
    • Spent Fuel Pool aging issues
    • I&C challenges
ceng pilot project ginna nmp1
CENG Pilot Project – Ginna & NMP1
  • Concrete
    • Tendon Monitoring: Installation of optical strain gages on 20 tendons (160 tendons total)
    • Digital Image Correlation
  • Reactor Metals
    • Baffle Bolts: 180 have been replaced at Ginna, 6 will be transferred to ORNL for evaluation
    • RPV specimens: three have been selected from Ginna
    • Top Guide Cracking
ii c issues present significant challenges for npps over extended lifetimes
Fleet-wide benefits of modern control technologies are high but the risks executing large scale modernization are also high because of regulatory and cost uncertainties II&C issues present significant challenges for NPPs over extended lifetimes
  • R& D Scope:
    • Next generation control room, IIC, and automation – pilot projects
    • Next generation NDE tools – detection and characterization of degradation precursors
    • On-line monitoring – wireless infrastructure and life cycle prognostics
new i c technologies
New I&C Technologies
  • Industrial engagement established: Duke, Entergy, South Texas Project (STP), APS, PG&E and more.
  • Reconfigurable Digital Simulation Laboratory: Full-scope, full-scale PWR plant model
  • Pilot Projects – Real Time Truth
    • New Alarm Technologies:
      • Prioritized alarm lists, Single-screen displays for all alarms, Mode sensitive (different alarms for different plant modes),
      • Automation of key plant functions that cause nuisance alarms, Filtering alarms to most important information for operator
    • Outage Control Center Paradigm Shift: OCC and work execution center modernization
      • Outages represent source of risk – Safety, Schedule adherence, risk of revenue loss;
      • Manpower serves as substitute for technology in today’s OCC (little real time data, informatics poor work environment, Radio and telephone communications)
    • Configuration Management:
      • System configuration management as a cause of forced outages & tech spec non-compliance
      • Development & demonstration of wireless technologies for non-permanent instrumentation and for data streaming
risk informed safety margin characterization rismc
Risk-Informed Safety Margin Characterization (RISMC)

Combining Probabilistic and Mechanistic Modeling to provide Integrated Quantification of Aleatory and Epistemic Uncertainty

Area 1: Develop R7 to simulate plant dynamics and analyze safety margin

  • System analysis (multiple threats)
  • Tightly couple multi-physics
  • Coverage of scenarios
  • Computational efficiency
  • Appropriate model fidelity

Area 2: Develop a risk-informed, simulation-driven methodology to drive R7 in analysis of plant safety margin

Area 3: Develop models of passive SSCs for application within R7, coupling phenomenology to SSC behavior

r7 a next generation of system analysis code to support risk informed safety decision making
R7 : a Next Generation of System Analysis Code to Support Risk-Informed Safety Decision Making

 Go beyond the current technology manifested by legacy codes (RELAP5, SAPHIRE) developed at INL and used broadly by the industry and regulatory evaluation

  • The R7 Project is a community effort, to develop new safety analysis methods, cultivate new safety culture, and train new generation of nuclear engineers

Overcome limitations of 1970s’ “divide and conquer” paradigm

Push envelope of M&S in all physical processes involved in plant safety

FY10-FY11 Goals:

Develop R7 code architecture and test version engine

Analyze a plant safety issue of importance for life extension

Demonstrate viability of using the R7 capability to construct the RISMC “safety case” on the selected life extension issue

Leverage on advances in applied math, computer and computational science

advanced lwr nuclear fuels
Advanced LWR Nuclear Fuels

Goals:

  • Develop and demonstrate very advanced fuels – accident resistant fuel (post Fukushima)
  • Improve the fundamental scientific understanding of fuel behavior
  • Develop predictive tools for advanced nuclear fuel performance
  • Speed implementation of new fuel technologies to industrial application
ceramic composite cladding sic development
Ceramic Composite Cladding (SiC) Development
  • Collaborative efforts with INL, EPRI, Westinghouse, MIT & ORNL
  • Fueled irradiations on-going at HFIR for ceramic matrix composite clad with UO2 and UN based on MOX testing samples
  • ATR irradiations planned for unfueled and fueled experiments in FY-11
  • Start of Halden Reactor Project irradiation planning in FY-11
  • MIT PWR coolant experiment planning in FY-11

SiCf/Sic

Zircaloy-4

major advanced lwr nuclear fuels pathway deliverables
Major Advanced LWR Nuclear Fuels Pathway Deliverables
  • 2015:
    • Initial lead test rod design with advanced fuel and planning
    • Rod testing planning/design with advanced fuel
    • Development of advanced fuel with multiple technologies
    • Rod irradiation with advanced fuel.
  • 2020:
    • Initial advanced fuel lead test assembly licensing
    • Reload testing planning/design with advanced fuel
    • Reload irradiation with advanced fuel.
  • 2025:
    • Initial advanced fuel reload design
    • Initial core reload with advanced fuel
    • Irradiation program for increased enrichment bundles
    • Irradiation program for increased exposure bundles.
thin film coating of zirc 4 outer surface
Thin Film Coating of Zirc-4 Outer Surface
  • Create zirconium cladding that is less susceptible to fretting failures and hydrogen reactions than conventional cladding
  • Lower risk technology than fuel SiC cladding technology

Zircaloy-4 tube for TiN coating

  • before polishing
  • cut and polished
  • mounted zircaloy-4 tube on heater of PLD chamber
  • TiN-coated tube.
samples annealed at 900 o c for 1 hour
Samples Annealed at 900oC for 1 hour

Samples under optical microscope before annealing: uncoated Zr-4 (left), 1mm TiN coated Zr-4 (right)

Samples under optical microscope after annealing: uncoated Zr-4 (left), 1mm TiN coated Zr-4 (right)

Surface morphology is changed for the bare substrate, while the sample coated with TiNshowed little or no change (no dilamination) in surface.

high temperature water corrosion testing
High Temperature Water Corrosion Testing
  • (1) Bare Zirc-4 for reference corrosion test.
  • (2) TiN/1-um.
  • (3) SiC/1-um.
  • (4) (50nm-TiN/50nm-AlN)/1-um.
  • (5) TiAlN/1-um.
  • (6) ZrN/1-um.
  • (7) AlN/1-um.

The coatings that survive corrosion testing will go into ATR for neutron irradiation testing

slide19
Fuel Performance Modeling: High-resolution model for fuel/clad behavior and response to a missing pellet surface condition

High resolution, 3D fuel performance calculation reveals the impact of a missing pellet surface on stress state, temperature profile, etc. (i.e. full nuclear fuel performance) as a function of fuel history.

Fuel Temperature

Von Mises

Effective

Stress

Clad Temperature

economics and efficiency improvement
Economics and Efficiency Improvement
  • Cooling Water Issues
    • Water issues are major part of the Nuclear Regulatory Commission’s (NRC’s) Environmental Review (Part 51).
    • Regulatory issues such as EPA’s 316(b) rule which may require retrofitting once-through cooling with cooling towers – EPRI estimated cost $32B. A potential “showstopper” for long term operation.
    • Water consumption with cooling towers & cooling tower performance
  • System Integration and Emergent Issues
  • Power Uprates
    • Assist NPP life cycle management
eight collaborative project descriptions
Eight Collaborative Project Descriptions
  • Issue request for proposals to seek “out of the box” strategies from outside the industry to address the consumptive use of water.
  • Peer review paper on cooling concepts for nuclear power plants (technical and economic tradeoffs of OTC vs. CCC; water consumption implications, and fish protection measures).
  • Template for exemptions from Phase 1, §316(b) requirement for CCC for new plants, especially on Great Lakes; use as basis for comments or response to forthcoming Phase II rule.
  • Wet cooling tower performance R&D.
  • Dry cooling tower R&D.
  • Support priority effort on DOE-NE/DOE-FE/DOE-OE water issue action plan from energy security standpoint; identify activities for DOE funding.
  • Develop methodology or framework for holistic yet flexible environmental impact assessment; incorporate data from local ecology/aquatic studies.
  • Field demonstrations of screen technologies, especially wedge-wire and fine mesh screens. 
economics and efficiency improvement1
Economics and Efficiency Improvement

Power Uprates

Spent

Fuel

Pool

Criticality

Margins

Containment

Performance

and Integrity

Modeling

Cross-Cutting with Advanced LWR Fuels Development

Adverse

Flow

Effects

Goal: Facilitate power uprates for operating LWRs – identify, investigate obstacles and resolve issues that inhibit potential power uprates in aging fleet

containment response modeling
Containment ResponseModeling
  • Thermal mixing, stratification and gas transport in containments are important phenomena for plant safety
  • Current plant system codes (RELAP5, etc.) and severe accident codes (SCDAP, MELCOR, etc.) have no or empirical lumped parameter models
  • Mechanistic models are required to analyze the containment TH phenomena such as suppression pool heatup behavior and hydrogen transport (e.g. Fukushima accident scenarios)
  • EEI Pathway is improving and developing the BMIX++ code, which is a highly efficient and mechanistic code to analyze thermal mixing and stratification in large pools of liquid and large dry containment space.
  • POOLEX at Finland was designed to study Finland BWR suppression pool heat-up behavior. Part of POOLEX experimental data have been simulated with GOTHIC and Fluent by KTH. Open available reports provided excellent data to validate BMIX++ code.

J. Laine and M. Puustinen, Thermal stratification experiments with the condensation pool test rig, NKS-117, 2006.

bmix simulation results
BMIX++ Simulation Results
  • GOTHIC and FLUENT results are according to H. Li and P. Kudinov, GOTHIC code simulation of thermal stratification in POOLEX facility, NKS-19, 2009.
  • BMIX++ is 1000 times faster than CFD simulations.
why is thermal mixing stratification important for containment performance
Why is thermal mixing/stratification important for containment performance?
  • The suppression pool surface temperature is one major factor to determine the containment pressure.
  • The suppression pool temperature distribution strongly affects available Net Positive Suction Head (NPSHa):
      • NPSHa = (Pnoncondensible + Pv(Tsurface) – Pv(Tsuction) ) / ρg + Hpool -Hpump –Hloss
      • The vapor pressure difference could mean about 2.8 m or 30% of NPSH required.
      • This effect had not been analyzed in the existing NPSH sensitivity analysis,* though this effect is far larger than any single effect included (Table 1, NRC document).
      • Thermal stratification tends to form after initial fast blowdown stage and intensifies with time, especially in long time accidents.
  • Much improvement in containment simulation is required to model the slow & long lasting phenomena that are very important (revealed by the Fukushima accidents) but were inadequately addressed or ignored by the existing codes.

* NRC draft guidance for use of containment accident pressure in determining the NPSH margin of ECCS and containment heat removal pumps.

conclusion
Conclusion
  • The existing fleet of nuclear power plants provide the majority of the Nation’s non-carbon emitting electrical generation
  • The continued operation of the existing fleet is in the National interest as a key strategy for meeting climate change and energy supply goals
  • Federal efforts are essential to stimulate and encourage industry efforts as well as to address the longer-term, high risk research that industry can not address
  • Sustained R&D on long-term LWR operations is needed to identify issues and develop the technical basis that supports industry efforts to relicense and power uprate plants for long-term operation