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Engineering Challenges in U.S. Nuclear Reactor Safety Jack Grobe, Associate Director for Engineering and Safety Systems Office of Nuclear Reactor Regulation Massachusetts Institute of Technology February 11, 2008 Agenda NRC Overview NRC Educational Support Activities NRR Overview

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Engineering challenges in u s nuclear reactor safety l.jpg

Engineering Challenges in U.S. Nuclear Reactor Safety

Jack Grobe, Associate Director for Engineering and Safety Systems

Office of Nuclear Reactor Regulation

Massachusetts Institute of Technology

February 11, 2008


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Agenda

  • NRC Overview

  • NRC Educational Support Activities

  • NRR Overview

  • Operating and New Reactors

  • Operating Reactor Trends and Technical Issues

  • Digital Instrumentation & Control Issues


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NRC Mission

License and regulate the Nation’s civilian use of byproduct, source, and special nuclear materials to ensure adequate protection of public health and safety, promote the common defense and security, and protect the environment.


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NRC Regulatory Functions

  • Establish standards, regulations and requirements

  • Issue licenses for nuclear facilities and users of nuclear materials

  • Inspect facilities and users of nuclear materials to ensure compliance with requirements


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NRC Organization

  • Two Major Programs

    • Nuclear Reactor Safety Program

      • Accounted for 74% of NRC’s costs in FY 2007

    • Nuclear Materials and Waste Safety Program

      • Accounted for 26% of NRC’s costs in FY 2007




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NRC Grant Opportunities

  • NRC FY 2008 Budget includes $ 19.7 million in educational support to colleges, universities, and trade schools.

  • NRC Grant Programs:

    • Nuclear Education Grant Program

      • $4.7 million towards curriculum development.

    • Scholarships and Fellowship Grant Program

      • $15 million towards scholarships and fellowships at colleges, universities, and trade schools.


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Nuclear Education Grant Program

  • Description

    • Provides funding support to institutions of higher education to support courses, studies, training, curricula, and disciplines pertaining to the NRC mission.

  • Process (FY08)

    • Applicants submit letter of intent – Jan. 22

    • NRC issues invitation for full proposal – Feb. 15

    • Full proposal submissions – March 21

  • Additional Information

    • www.nrc.gov/about-nrc/grants.html


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Scholarship and Fellowship Grant Program

  • Description

    • Allows students to pursue an education in science, engineering, or other fields or study related to the NRC mission.

  • Requirements

    • Recipients agree to 1 year employment for each full or partial year of academic support.

    • Eligible applicants attend regionally accredited 4-year U.S. public and private institutions of higher education.

    • Strong interest in areas of probabilistic risk, thermodynamics, mechanics, fuels, and digital instrumentation and controls.

  • Additional Information

    • Solicitation will be posted on Grants.gov during the week of February 11, 2008.


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Office of Nuclear Reactor Regulation

  • What does NRR do?

    • Contributes to the NRC mission to protect public health, safety, and the environment for reactors regulated under 10 CFR Part 50

  • How does NRR do this?

    • By developing and implementing the following major reactor programs:

      - Rulemaking

      - Licensing

      - Oversight, and

      - Incident Response




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EPR

Amarillo Power

Ameren UE

PPL Generation

UNISTAR

AP1000

Duke

NuStart

Progress Energy

S.C. Electric & Gas

Southern Co.

US APWR

TXU Power

US ABWR

NRG Energy

ESBWR

Dominion

Entergy

NuStart

Potential New Reactor Applicants


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Recent NRR Activities

  • Safety of Operating Reactors

    • Recent Trends

  • Operating Experience

    • Safety Culture

    • Extended Power Uprates (EPU) – Steam Dryer Analysis

    • Alloy 82/182 Dissimilar Metal Butt Weld (DMBW)



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2006 Average Industry Performance(Secy-07-0063 / 4-3-07 / ML070660099)


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2006 Average Industry Performance(Secy-07-0063 / 4-3-07 / ML070660099)


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2006 Average Industry Performance(Secy-07-0063 / 4-3-07 / ML070660099)



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Safety Culture

  • Human performance and organization effectiveness are the keys to sustain safe operations

  • Many complex factors influence successful performance

  • Several facilities exhibited weak performance

    • Peach Bottom

    • Millstone

    • D.C. Cook

    • Clinton

    • Davis-Besse

    • Palo Verde

  • INPO and NRC engaging in organizational effectiveness assessment


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Safety Culture

  • Fall 2002 – INPO SOER

    • Immediate requirement

    • All facilities perform assessment

    • Report results to INPO

  • IAEA Documents

    • INSAG-15, “Key Practical Issues in Strengthening Safety Culture.”

      http://www-pub.iaea.org/MTCD/publications/PDF/Pub1137_scr.pdf

    • INSAG-4, “Safety Culture.”

      http://www-pub.iaea.org/MTCD/publications/PDF/Pub882_web.pdf

    • INSAG-13, “Management of Operational Safety in Nuclear Power Plants.”

      http://www-pub.iaea.org/MTCD/publications/PDF/P083_scr.pdf

  • NRC Regulatory Information Summary 2006-13, “Information on the Changes Made to the Reactor Oversight Process to More Fully Address Safety Culture”

    • Evaluate safety culture weaknesses

    • Structured process for assessing results

    • Cross cutting attributes

      http://www.nrc.gov/reading-rm/doc-collections/gen-comm/reg-issues/2006/ri200613.pdf


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Extended Power Uprates- Steam Dryers

  • Many licensees have been implementing power uprates to increase nuclear power electric output.

  • Some boiling water reactors have experienced adverse flow effects on safety related and non-safety related equipment during power uprate operation.

  • Most significant problems occurred at Quad Cities Units 1 and 2, which experienced failures of steam dryers and electromatic relief valves during Extended Power Uprate (EPU) operation.

  • Licensees need to consider potential adverse flow effects when planning to implement a proposed power uprate.



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QC1 Steam Dryer FailureNovember 2003(close-up)

Missing portion of outer bank vertical plate, approx. 6 in. x 9 in.


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QC2 Steam Dryer FailureMarch 2004

Tie bar to attachment welds

Plate attachment stitch weld

Tip of gusset plate




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ASB Modification Compared to Pre-ASB Data

Relief

valve

ASB


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PWR Alloy 82/182 Dissimilar Metal Butt Welds Compared to Pre-ASB Data

Surge Nozzle-to-Safe End

Dissimilar Metal Weld

Safety/Relief Nozzle-to-Safe End Dissimilar Metal Weld


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Alloy 82/182 Dissimilar Compared to Pre-ASB DataMetal Butt Welds

  • Stress corrosion cracking

    • Residual stress

    • Challenging environment

    • Susceptible material

  • NDE identified flaws in Alloy 82/182 welds due to PWSCC beginning in 2000.

  • Circumferential involvement was limited until 2006.

  • 5 circumferential indications discovered in Wolf Creek pressurizer nozzle welds in October 2006:

    • Up to 50 percent circumferentially; 40 percent through-wall

    • One weld contained three circumferential cracks

      NRC concerns:

    • First case of multiple, long, circumferential flaws


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NRC Wolf Creek Flaw Evaluation Scoping Study Compared to Pre-ASB Data

  • NRC staff evaluated integrity of pressurizer nozzles using ASME Section XI flaw evaluation methodologies

    • Result of study showed little margin between leak and rupture for nozzle with the largest flaw

  • NRC staff obtained agreements from all licensees to resolve issue in 2007.

  • Industry concern with nine units that did not have scheduled outages in 2007.

  • NRC staff agreed that industry advanced finite element analyses could be conducted to address staff concern.


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Advanced Finite Element Analyses Compared to Pre-ASB Data

  • Industry’s advanced finite element analyses were completed in August 2007 (ML072410240)

  • NRC staff completed confirmatory analyses (ML072470394)

  • NRC staff issued safety assessment (ML072400199) concluding reasonable assurance that PWSCC in pressurizer nozzles of the nine plants would not lead to rupture without adequate time to shutdown the facility

    • allowed operation to the spring 2008 outages to complete pressurizer nozzle inspections




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Advanced FE Analysis Crack Growth Compared to Pre-ASB Data

36


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Technical Issues: Operating Compared to Pre-ASB DataReactors

  • Fire Protection

  • Proactive materials degradation assessment

  • Integrated digital I&C and human- machine interface


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Fire Protection Compared to Pre-ASB Data

  • Fire barrier performance

    • Potential issues with performance of Thermo-Lag and other barrier systems

    • Confirmatory studies

      • HEMYC performance

  • Cable performance

    • Potential for multi-conductor cable failures in fires

    • Confirmatory studies

      • CAROLFIRE tests


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Technical Issue: Materials Degradation Compared to Pre-ASB Data

  • Materials degradation has been experienced worldwide since the inception of nuclear power plant operation

  • Degradation is expected to continue as plants age

  • Develop technical basis for proactive materials degradation management

    • Identify components where

      degradation can reasonably

      be expected in the future

    • Coordinate research for

      effective implementation

      of proactive materials

      degradation management


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Digital I&C Issues Compared to Pre-ASB Data


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Current Technology Compared to Pre-ASB Data


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Future Technology Compared to Pre-ASB Data


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Digital I&C Challenges Compared to Pre-ASB Data

  • Increased complexity

    • Consolidation of discrete analog functions into single digital system

    • Potential consolidation of independent safety systems into a single digital system

    • Potential new failure modes

  • Limited operational history in nuclear applications

  • Specialized staff skills


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Digital I&C Project Compared to Pre-ASB Data

  • Steering Committee established January 2007

  • Seven Task Working Groups (TWGs) have been formed:

    • Cyber Security

    • Diversity and Defense-In-Depth

    • Risk-Informed Digital I&C

    • Highly Integrated Control Room – Communications

    • Highly Integrated Control Room – Human Factors Issues

    • Licensing Process Issues

    • Fuel Cycle Facilities

  • Goal is to develop Interim Staff Guidance (ISG) licensing guidance for digital issues

  • Developed Digital I&C Website:

    http://www.nrc.gov/about-nrc/regulatory/research/digital.html


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Cyber Security ISG Compared to Pre-ASB Data

  • Clarifies the NRC staff’s guidance on cyber security requirements

  • Addresses entire plant including safety and nonsafety systems programmatically

  • Reg. Guide 1.152 positions 2.1 through 2.9 address cyber security licensing criteria for safety systems

  • ISG issued and placed on Digital I&C Website on December 31, 2007


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Diversity and Defense-in-Depth (D3) ISG Compared to Pre-ASB Data

  • Clarifies what is adequate diversity

  • Clarifies applicability of Common Cause Failure (CCF) in digital systems

  • CCF even though beyond design basis requires diverse RPS (RTS and ESFAS) for digital systems

  • Clarifies time required (30 minutes) for taking credit for manual actions

  • Clarifies position on system level vs. component level actuation

  • ISG issued and placed on Digital I&C Website on 9/26/2007


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Risk Informing Digital I&C ISG Compared to Pre-ASB Data

  • Guidance for use of PRA evaluations

  • Addresses the appropriateness of risk-insights to resolve issues

  • Establishes an acceptable state-of-the-are model methodology

  • ISG scheduled to be issued on 3/28/08


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Highly Integrated Control Room - Communications ISG Compared to Pre-ASB Data

  • Addresses communication between different safety divisions and between safety division and non-safety equipment

  • Provides guidance for command prioritization

  • Provides guidance for Multidivisional Control and Display Stations

  • ISG was issued on 9/28/2007 and placed on the Digital I&C Webpage


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Highly Integrated Control Room – Human Factors Issues ISG Compared to Pre-ASB Data

  • Clarifies approach the NRC staff would use to evaluate acceptable human factors aspects of highly-integrated control rooms

  • Guidance on minimum inventory for the Main Control Room and the Remote Shutdown Facility

  • Guidance on development, use, update (including software controls), and automation of Computer-Based Procedures System

  • ISG was issued on 9/28/2007 and placed on the Digital I&C Webpage


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Licensing Process Issues ISG Compared to Pre-ASB Data

  • Clarifies the level of detail necessary for licensing reviews

  • Guidance on applicability of SRP, Ch. 7

  • Clarifies protocols for licensing action application development

  • Guidance on cyber security

  • ISG (w/o cyber security) scheduled to be issued on 7/31/08


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Fuel Cycle Facility ISG Compared to Pre-ASB Data

  • Guidance for reviewing adequacy of cyber security protective measures

  • Clarifies adequate diversity and defense-in-depth design features

  • Guidance on channel independence for criticality and non-criticality related safety actions

  • Guidance on separation of safety-related functions from non-safety related functions in common operator interface devices

  • Clarifies acceptable use of software for safety functions to minimize common cause failures

  • ISG scheduled to be issued on 10/31/08


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Path Forward for Digital I&C Compared to Pre-ASB Data

  • After issuance of ISGs, industry standards will be updated to include the applicable portions of the ISGs as part of the standards

  • NRC staff will issue Regulatory Guidance to endorse the applicable industry standards

  • NRC staff will issues updated/new Standard Review Plans to address the updated Regulatory Guides

  • The ISGs will be cancelled on the completion of the above items


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Questions ? Compared to Pre-ASB Data