- About NRC           - The Commission
      - Statutory Authority
      - Strategic Plan
      - Budget & Performance
      - Organization & Functions
      - Locations
      - History
      - Values
      - Policymaking
      - Radiation Protection
      - Fire Protection
      - Regulation
      - Emergency Preparedness   and Response
      - Public Affairs
      - Congressional Affairs
      - International Programs
      - State and Tribal Programs
      - Alternative Dispute   Resolution Programs
      - Contact Us
      - Employment
      - Contracting
      - Grants
      - Civil Rights
      -     
    - NuclearReactors           - Power Reactors
      - Research and Test Reactors
      - Operating Reactors
      - Operator Licensing
      - New Reactors
      - Advanced Reactors
      - Operator Licensing for New Reactors
      - Specific Plants of Interest
      -     
    - NuclearMaterials           - Special Nuclear Material
      - Source Material
      - Byproduct Material
      - Med, Academic & Ind Uses
      - Source Materials Facilities
      - Uranium Recovery Facilites
      - Fuel Cycle Facilities
      - Materials Transportation
      -     
    - RadioactiveWaste           - Low-Level Waste
      - Waste Incidental to Reprocessing
      - High-Level Waste
      - Uranium Mill Tailings
      - LLW Disposal
      - HLW Disposal
      - Storage of Spent Nuclear Fuel
      - Transporation of Spent Nuclear Fuel
      -     
    - NuclearSecurity           - Domestic Safeguards
      - Information Security
      - Radioactive Material
      - &;   
    - Public Meetings& Involvement           - Open Government
      - Public Meeting Schedule
      - Commission Schedule
      - Conferences & Symposia
      - Documents for Comment
      - Involving Stakeholders
      - Info Quality Guidelines
      - Subscribe to E-mail Notices 
      -     

UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001

December 11, 1997

**Addressees**

All holders of operating licenses for nuclear power reactors except those who have permanently ceased operations and havecertified that fuel has been permanently removed from the reactor vessel.

**Purpose**

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to potential genericproblems related to the occurrence and prediction of flow-accelerated corrosion (FAC) in extraction steam systems. It isexpected that recipients will review the information for applicability to their facilities and consider actions, as appropriate,to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore,no specific action or written response is required.

**Description of Circumstances**

On April 21, 1997, Omaha Public Power District's Fort Calhoun Station, while operating at 100-percent power, experiencedan approximate 0.56 m2 [6 ft2] rupture of a 30.5-centimeter [12-inch]-diameter sweep elbow (radius equal to five times thepipe diameter) in the fourth-stage extraction steam piping. The operator, upon hearing steam noise and observing steamrising from the turbine deck, believed that a steam line had broken and manually scrammed the reactor. As a precaution,emergency boration was initiated. The main turbine tripped automatically as a result of the reactor trip. The turbine trip hadthe effect of isolating the rupture. Plant systems and related parameters responded as expected during the event.

The steam line rupture damaged a nonsafety-related electrical load center in the vicinity of the pipe break. Additionally,collateral damage was experienced in several cable trays and pipe hangers, and insulation containing asbestos was blownthroughout the turbine building. Certain portions of the fire protection system actuated in response to fusible links in thesprinkler heads melting because of high temperature. Because there were no personnel in the immediate vicinity of therupture, no one was injured.

The fourth-stage extraction steam system emanates from the outlet of the high-pressure turbine and preheats the feedwaterheaters. The design operating conditions in the piping are 2068 kilopascal gauge [300 psig] and 218C [425F], with asteam quality of approximately 92 percent. The piping is fabricated of A-106B carbon steel and has a nominal wallthickness of 0.953 centimeter [0.375 inch]. The licensee's root cause assessment attributed the failure to FAC in theextraction steam piping. Initial indications of degradation in the extraction steam line at the Fort Calhoun facility were firstdiscovered in 1985, when the furthest upstream long-radius elbow (radius equal to one and a half times the pipe diameter)was replaced because of a pinhole leak. At that time the next upstream sweep elbow was also replaced.

**Discussion**

The fourth-stage extraction steam system had been recognized as a system that was susceptible to erosion and/or corrosion.It was, therefore, being monitored by the licensee's erosion and corrosion control program. Part of the licensee program wasutilizing the CHECWORKS computer code to identify high-wear-rate areas to be selected for inspection.

The CHECWORKS model for the fourth-stage extraction steam piping predicted that long-radius elbows would wear at ahigher rate than the sweep elbows when exposed to similar conditions. Using CHECWORKS predictions, the licenseeinspected and replaced all four long-radius elbows, but the failed sweep elbow was never inspected.

Part of the licensee's corrective actions following the rupture included inspecting all sweep elbows that had not beenpreviously inspected. The measured wall thickness (0.112 centimeter [0.044 inch]) of the furthest downstream sweep elbowin the fourth-stage extraction piping was also significantly below the minimum wall thickness (0.272 centimeter [0.107inch]) specified by code requirements and had to be replaced. Additionally, another sweep elbow in the fourth-stageextraction piping was also replaced because the wear (measured wall thickness of 0.394 centimeter [0.155 inch]) wasconsidered excessive, even though it was not below the minimum allowable thickness.

The CHECWORKS predictions of the wear in the fourth-stage extraction steam system were not consistent with the actualobserved wear rates as measured on the components, that is, sweep elbows showed substantially greater wear thanpredicted.

Subsequent investigations by the licensee determined that the inconsistencies between predicted and actual wear were dueto two factors. First, the "line correction factor" calculated by CHECWORKS for the fourth-stage extraction steamline wasnot within the acceptable range specified in the CHECWORKS users' manual. The line correction factor in CHECWORKSis used to adjust wear rate predictions in a given line to account for plant operating conditions that may vary with time. It isdetermined by comparing predicted wear to measured wear at locations in the line which have been inspected.

In order for the CHECWORKS predicted wear rate for a location to be valid, the line correction factor must be between 0.5and 2.5. For the line containing the sweep elbow that failed, the wear rates calculated by CHECWORKS used a linecorrection factor that was outside this range. Therefore, as specified in the CHECWORKS users' manual, the predictedwear rates for this line were not valid.

Second, the line correction factor was biased and thus underpredicted the wear rates. In 1987, the licensee updated theparameters used by CHECWORKS to include the wear measured in a long-radius elbow. One of the inputs inCHECWORKS is the length of time the component has been in service. The licensee assumed that this elbow had been inservice since initial operation of the plant in 1973; however, the elbow had been replaced in 1985. Therefore, the actualwear occurred over 2 years rather than the presumed 14 years. Thus, the period of time that was assumed for the wear tohave occurred caused CHECWORKS to calculate a line correction factor that underestimated the wear rates for sites in theline.

This event revealed the importance of understanding the limitation of methodologies used in computer programs and ofincorporating accurate plant-specific data from nondestructive examination programs.

**Related Generic Communications**

In NRC Bulletin 87-01, "Thinning of Pipe Walls in Nuclear Power Plants," July 9, 1987, the staff asked licensees andapplicants to inform the NRC about their programs for monitoring the wall thickness of carbon steel piping (Accession No.8707020018).

In NRC Generic Letter 89-08, "Erosion/Corrosion Induced Pipe Wall Thinning," May 2, 1989, the staff asked licensees andapplicants to implement long-term erosion/corrosion monitoring programs (Accession No. 8905040276).

Additionally, the following NRC information notices (INs) provide information about similar events related to FAC:

This information notice requires no specific action or written response. However, recipients are reminded that they arerequired to consider industry-wide operating experience (including NRC information notices) where practical, when settinggoals and performing periodic evaluations under Section 50.65, "Requirement for monitoring the effectiveness ofmaintenance at nuclear power plants," to Part 50 of Title 10 of the Code of Federal Regulations. If you have any questionsabout the information in this notice, please contact the technical contact listed below or the appropriate Office of NuclearReactor Regulation (NRR) project manager.

----