Information Notice No. 96-11: Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations

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

                               February 14, 1996

                           MECHANISM PENETRATIONS 


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


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the increased likelihood of stress corrosion
cracking of pressurized water reactor (PWR) control rod drive mechanism (CRDM)
penetrations if demineralizer resins contaminate the reactor coolant system
(RCS).  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 supplement are not NRC requirements; therefore, no specific action or
written response is required.


In 1990, the NRC staff issued Information Notice 90-10, "Primary Water Stress
Corrosion Cracking (PWSCC) of Inconel 600," informing PWR licensees that PWSCC
was an emerging technical issue.  PWSCC was noted in Inconel 600 pressurizer
heater sleeve penetrations at a domestic PWR facility.  The NRC staff
determined that the safety significance of the cracking was low because the
cracks were axial, had a low growth rate, and were in a material with an
extremely high flaw tolerance (high fracture toughness).  Accordingly, the
cracks were unlikely to propagate very far.  

In December 1991, after cracks were found in a CRDM penetration in the reactor
head at a French plant, an NRC action plan was implemented to address PWSCC at
all U.S. PWRs.  The NRC staff met with the Westinghouse Owners Group, the
Babcock and Wilcox Owners Group, and the Combustion Engineering Owners Group
to discuss their respective programs for investigating PWSCC of Inconel 600
and to assess the possibility of cracking of CRDM penetrations in their
respective plants.  Subsequently, the staff asked the Nuclear Management and
Resources Council, now the Nuclear Energy Institute, to coordinate future
industry actions because the issue was applicable to all PWRs.  Each owners 

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group submitted individual safety assessments, dated February 1993, through
Nuclear Energy Institute to the NRC on the CRDM penetration cracking issue. 
In July 1993, the Institute submitted to the NRC proposed acceptance criteria
for flaws identified during inservice examination of CRDM penetrations.  On 
the basis of the owners group analyses and the European experience, the NRC
staff concluded, in a safety assessment dated November 19, 1993, (NRC
Accession No. 9403020162), that there is a high probability that CRDM
penetrations at U.S. plants may contain similar axial cracks caused by PWSCC. 

The Electric Power Research Institute is engaging in ongoing research on
methods for mitigating PWSCC.  They also have developed a demonstration
program to ensure that inspections performed on CRDM penetrations are highly
reliable in detecting and determining the size of flaws.

The first of three U.S. inspections took place in the spring of 1994 at the
Point Beach Nuclear Generating Station.  No indications were uncovered in the
CRDM penetrations.  The eddy current inspection at the Oconee Nuclear Station,
Unit 2, in the fall of 1994 revealed 20 indications in one penetration. 
Ultrasonic testing did not reveal the depth of these indications because they
were shallow.  These indications may be associated with the original
fabrication and may not grow; however, the licensee has committed to reexamine
this penetration during the next refueling outage.  An examination of the CRDM
penetrations at the Donald C. Cook Nuclear Plant Unit 2 in the fall of 1994
revealed three clustered indications in one penetration.  The indications were
46 mm (1.7 in.), 16 mm (0.63 in.), and 7 mm (0.28 in.) in length and the
deepest flaw was 6.8 mm (0.27 in.) deep.  The tip of the 46 mm (1.7 in.) flaw
was just below the J-groove weld.  These results are consistent with the PWR
owners group analyses, the NRC staff safety evaluation of the owners group
analyses, and the PWSCC found in the CRDM penetrations in European reactors. 
The results of these inspections are documented in Safety Evaluation Reports
dated January 1995 for the D.C. Cook Plant (Accession Nos. 9504050173,
9504050168, 9503220149) and January 1995 for the Oconee Plant (Accession 
No. 9503270178).  

Description of Circumstances

Early in 1994, an inspection for PWSCC at a reactor in Spain identified cracks
which were apparently initiated by high sulfate levels in the reactor coolant
system.  Two cation resin ingress events had occurred at the reactor.  In
August 1980, 40 liters of cation resin entered the coolant system.  In
September 1981, a mixed-bed demineralizer screen failed and five to eight
times as much resin entered the coolant system as that entering in the August
1980 event.  The coolant conductivity remained high for at least 4 months
after the ingress.  The increase in conductivity was attributed to acid .                                                          IN 96-11
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sulfate.  Sulfates were found around the crack areas and on the fracture
surfaces.  It is important to note that sulfate cracking occurs in lower
stress regions than does PWSCC.  The Spanish reactor has 37 CRDM penetrations,
of which 20 are active and 17 are spare.  Of the 17 spare penetrations, 16
showed stress corrosion cracking and intergranular corrosion.  The cracks were
both axial and circumferential.  Four of the active CRDM penetrations had
significant axial and circumferential cracking.  

Westinghouse notified the Westinghouse Owners Group plants, the Babcock and
Wilcox Owners Group plants, and the Combustion Engineering Owners Group plants
of the Spanish reactor incident by issuing NSAL-94-028.  Westinghouse informed
the NRC staff, during a public meeting on August 24, 1995, that NSAL-94-028
recommends that PWR licensees review their primary coolant system water
chemistry to verify that they have not had significant primary system resin
bed intrusions, and that U.S. PWRs review their RCS chemistry and other
operating records relative to sulfur ingress events.  Westinghouse also
reported during this meeting that no other plant had been found worldwide that
has experienced cracking similar to that at the Spanish reactor and that the
U.S. plant inspection results agreed in general with the worldwide experience.
The Westinghouse staff further reported that U.S. plants routinely monitor RCS
conductivity, follow the Electric Power Research Institute guidelines on
primary water chemistry, and monitor for sulfates three times a week. 
Westinghouse concluded that no immediate safety issue exists and that the
conclusions in its CRDM safety evaluation, dated February 1993 (WCAP-13565,
NRC Accession No. 9312090177), remain valid.  


The NRC staff is not aware of any significant primary system resin bed
intrusions at any U.S. PWR.  However, if any significant resin intrusions have
occurred at U.S. PWRs, residual stresses are likely sufficient to cause
circumferential intergranular stress corrosion cracking.  The NRC staff has
agreed to meet with National Electric Institute and the PWR owners groups in
early 1996 to continue discussions on this issue.

On the basis of the results of the inspections at three U.S. PWRs, the NRC
staff continues to conclude, as stated in the 1993 safety evaluation, that
there is a high probability that CRDM penetrations at other plants may contain
similar axial cracks caused by PWSCC.  

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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
the technical contacts listed below.

                                           signed by B.K. Grimes

                                      Dennis M. Crutchfield, Director
                                      Division of Reactor Program Management
                                      Office of Nuclear Reactor Regulation

Technical Contacts:  Keith A. Wichman, NRR                
                     (301) 415-2757             

                     James A. Davis, NRR
                     (301) 415-2713

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