United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 93-21: Summary of NRC Staff Observations Compiled During Engineering Audits or Inspections of Licensee Erosion/Corrosion Programs

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D.C.  20555

                                March 25, 1993

                               LICENSEE EROSION/CORROSION PROGRAMS


All holders of operating licenses or construction permits for light water
nuclear power reactors.


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to inform addressees of NRC observations of the design and
implementation of licensee erosion/corrosion programs.  These observations are
derived from a number of audits and reactive inspections performed by NRC
audit teams and by Region I office inspectors.  It is expected that recipients
will review the information for applicability to their facilities.  However,
suggestions contained in this information notice are not NRC requirements;
therefore, no specific action or written response is required.


Erosion/corrosion has occurred in systems containing carbon steel components
in certain nuclear plant systems, such as main feedwater or other power
conversion systems, which are important to the safe operation of the plant.  
Failures at the Millstone Nuclear Power Station, Units 2 and 3, and numerous
repairs and replacements in extraction steam and moisture separator systems at
other facilities indicate that erosion/corrosion continues to be a problem in
balance of plant systems.  Some boiling water reactor (BWR) licensees have
discovered erosion/corrosion in American Society of Mechanical Engineers
(ASME) Code Class 1 portions of their feedwater systems.  Some pressurized
water reactor (PWR) licensees have discovered erosion/corrosion in ASME Code
Class 2 portions of their feedwater systems and in ASME Code Class 3 portions
of their main steam systems.  The worn areas affected by erosion/corrosion in
Class 1 piping have typically been in regions directly downstream of feedwater
reducing tees which branch to the feedwater risers.  The worn areas of the
Class 2 feedwater piping have typically been just upstream of the feedwater
nozzles to the steam generator, although the licensee for the Diablo Canyon,
Unit 1, nuclear power plant recently reported erosion/corrosion wear behind a 


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thermal sleeve on the interior of the feedwater nozzle and on the feedwater
nozzle itself.  The worn area in the Class 3 system at the North Anna, Unit 1,
nuclear power plant was located in a main steam drain line upstream of a main
steam trap valve (an automatic isolation valve).  These findings indicate that
erosion/corrosion can occur in safety-related portions of plant feedwater and
main steam systems.  

Summary of NRC Observations from Audits and Inspections 

The results of recent NRR audits and regional inspections of licensee
erosion/corrosion programs indicate that most licensees have spent substantial
time and resources to implement erosion/corrosion programs in accordance with
the guidelines of Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall
Thinning."  This information notice presents some specific observations on the
design and implementation of erosion/corrosion programs, which vary in scope. 

Most of the problems that licensees have had in implementing erosion/corrosion
programs pertained to weaknesses or errors in (1) using predictive models, 
(2) calculating minimum wall thickness acceptance criteria of the code, 
(3) analyzing the results of ultrasonic testing (UT) examinations, (4) self-
assessment of erosion/corrosion program activities, (5) dispositioning
components after reviewing the results of the inspection analyses, or 
(6) repairing or replacing components that failed to meet the minimum wall
thickness acceptance criteria of the licensee.

Concerning item (1) above, some licensees have selected systems and components
for UT examinations based on the analytical results of predictive models, such
as the CHEC or CHECMATE computer codes of the Electric Power Research
Institute or the Massachusetts Institute of Technology method described in
NUREG/CR-5007, "Prediction and Mitigation of Erosion-Corrosive Wear in
Secondary Piping Systems of Nuclear Power Plants," which ranks systems and
components according to their erosion/corrosion susceptibility.  Other
licensees have selected systems and components based on engineering judgment. 
Recent events at Millstone 2 [Licensee Event Report (LER) 50-336/91-12] and
Millstone 3 (LER 50-309/92-07), before the licensee restructured its
erosion/corrosion program, and at Maine Yankee (LER 92-007-00), indicate that
such erosion/corrosion programs based on engineering judgment alone may lack
the scope needed to predict these phenomena in high energy, carbon steel

The Office of Nuclear Reactor Regulation (NRR) audit teams and regional
inspection teams found, for those cases in which CHECMATE was used as a
predictive model, that licensees sometimes made errors in entering the proper
plant parameters into the computer code.  The most common of these errors had
to do with selecting the proper geometry code for a component of a system. 
Errors in entering plant parameters into CHECMATE can result in errors in the
predicted wear rate.

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Concerning item (4) above, the inspection teams also observed that some
licensees do not have programs to assess their own erosion/corrosion-related
activities, especially for activities associated with computer-generated
analyses.  Licensee Quality Assurance (QA) programs typically did not cover
the implementation of the erosion/corrosion programs for safety-related
piping, although some licensees assessed their erosion/corrosion programs
during audits of their inservice inspection programs.  ASME Section XI,
referenced in 10 CFR 50/55a(g)(4), requires that activities within the
jurisdiction of the Code be covered by a quality assurance program (ASME
Section XI, IWA 1400).

Concerning item (6) above, the inspections also revealed that not all repairs
or replacements of erosion/corrosion-worn safety-related components are being
performed in accordance with the applicable requirements of the ASME Code,
Section XI, Article IWA 4000/7000.  One licensee (Susquehanna; Combined
Inspection Reports 50-387/92-05 and 50-388/92-05) repaired a worn, Class 1
feedwater riser by depositing a weld overlay on the exterior of the pipe. 
Repairs or replacements of worn components in systems designed or reclassified
as ASME Code Class 1, 2, or 3 systems must satisfy the requirements of 
10 CFR 50.55a(g) and ASME Code, Section XI.  Reinforcement of worn areas by
weld overlays on the outside of high energy, safety-related pipe do not comply
with the ASME Code because the Code requires that flaws be removed from the
pipe before the weld material is applied (ASME XI, IWA-4300) for repairs of
safety-related pipe.

NRC inspection teams have also observed the following during inspections of
licensee erosion/corrosion programs: 

o    An improper determination of code minimum wall thickness acceptance
     criteria, resulting in improper disposition of degraded components
     (Salem Unit 1; Inspection Report 50-272/92-08) 

o    A lack of baseline thickness measurements (history) on originally
     designed piping (Hope Creek; Inspection Report 50-354/92-11) and on
     replacement piping before the replacement piping is put into service
     (Millstone Unit 1; Inspection Report 50-245/92-80)

o    Inconsistency in reproducing UT grid locations during later UT
     examinations of the same component (Maine Yankee; Inspection 
     Report 50-309/92-14)

o    Use of engineering personnel who are not familiar with plant operating
     conditions, plant as-built designs, or erosion/corrosion history
     (Millstone Unit 2; Inspection Report 50-336/91-81 and Hope Creek;
     Inspection Report 50-354/92-11) 

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This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.

                                      ORIGINAL SIGNED BY

                                   Brian K. Grimes, Director
                                   Division of Operating Reactor Support
                                   Office of Nuclear Reactor Regulation

Technical contacts:  Jim Medoff, NRR
                     (301) 504-2715

                     Krys Parczewski, NRR
                     (301) 504-2705

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