ML20056C660
| ML20056C660 | |
| Person / Time | |
|---|---|
| Issue date: | 05/12/1993 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Rogers K, Selin I, The Chairman NRC COMMISSION (OCM) |
| References | |
| NUDOCS 9307130218 | |
| Download: ML20056C660 (4) | |
Text
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May 12, 1993
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MEMORANDUM FOR: The Chairman N
Comissioner Rogers Comissioner Curtiss Comissioner Remick Comissioner de Planque FROM:
James M. Taylor Executive Director for Operations
SUBJECT:
STATUS REPORT ON PRIMARY WATER STRESS CORROSION CRACKING OF ALLOY 600 COMPONENTS Enclosed for your information is an update on primary water stress corrosion cracking of Alloy 600 (INCONEL) in the reactor coolant pressure boundary. The staff will send an information paper to the Comission after it issues the safety evaluation on this subject. The staff is scheduled to issue its safety evaluation 60 days after it receives the two remaining safety analyses of the consequences of postulated cracking of the control rod drive mechanism (CRDM) penetrations from the PWR Owners Groups at the end of June 1993.
3 Having reviewed the information to date, including the inspection results and findings, the staff maintains its view that this issue is of low safety significance since all cracks reported to date, with perhaps one exception, are short in length and axially oriented in an extremely flaw-tolerant material. Further, the effects of wastage by borated water of a creviced area such as between CRDM penetration and the reactor head have been evaluated on the bases of laboratory testing and similar field experience. The results indicate that any degradation would occur very slowly and, therefore, an event such as ejection of a CRDM continues to be unlikely.
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'3 There are safety concerns related to minimizing worker exposure associated with inspections. Currently the U.S. industry is developing remotely operated n
. inspection equipment, criteria for repair, and remotely operated tools to implement repairs, if needed. Based on the unlikely failure of a CRDM penetration and low safety significance of CRDM leakage, we believe that there is sufficient time available to implement a well-thought-out and well-planned inspection, evaluation and repair program.
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James M. Taylor James M. Taylor Executive Director for Operations
Enclosure:
As stated cc:
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ENCLOSURE m~
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STATUS REPORT ON PRIMARY WATER STRESS CORROSION CRACXING OF ALLOY 600 (Inconel) COMPONENTS 3G7 '
Following a 1989 Teakage from an Alloy 600 pressurizer heater sleeve penetration at Calvert Cliffs Unit 2, primary water stress corrosion cracking was identified to the Commission as an emerging issue.
In addition to leaks at the heater sleeve, other leaks have been occurring since 1986 in several Alloy 600 pressurizer instrument nozzles at both domestic and foreign reactors from several different nuclear steam supplier vendors. Since 1989, the staff has been actively pursuing this issue with the Combustion Engineering Owners Group (CEOG) and the Babcock & Wilcox Owners Group (B&WOG). The Westinghouse Owners Group (WOG) was contacted at that time and responded that it did not use Alloy 600 for penetrations or nozzles in the pressurizers.
Initially, the CEOG and, about a year later, the B&WOG embarked on programs to: (1) locate areas where Alloy 600 was being used; (2) assess material susr.eptibility based on chemistry, heat treatment, and fabrication practices; (3) develop repair / replacement methods; and (4) determine inspection needs.
In 1991, a leak was discovered during a hydrostatic test in an Alloy 600 control rod drive mechanism (CRDM) penetration at the Bugey 3 plant in France.
At a meeting with the WOG on January 7,1992, the staff discussed the CRDM leak at the foreign plant and its relationship to domestic Westinghouse plants. WOG informed the staff that a program had been initiated in December 1991 to: (1) determine the root cause of the CRDM penetration cracking; (2) analyze the stress distributions in the CRDM penetrations of a typical domestic plant; (3) compare the design and operational characteristics of U.S.
and French plants to determine the likelihood for cracking; and (4) identify the need for additional efforts. The staff also met with CEOG and B&WOG on March 25 and May 12, 1992, respectively. Subsequently, the Nuclear Management and Resources Council (NUMARC) was asked to coordinate the PWR Owners Group efforts. More meetings were held on August 18 and November 20, 1992 and March 3, 1993.
At the March 3, 1993 meeting, NUMARC reported that CRDM penetrations have been inspected at 35 plants in France, Sweden, Switzerland, Japan, and Belgium. Of 1,174 penetrations, 46 had short, axially oriented flaws. Although additional minor flaws were found in the metallographic examination of the T-54 penetration removed from Bugey 3, sufficient information is not available to assess the results.
After reviewing the information to date and studying the inspection reports, the staff is unchanged in its view that this issue is of low safety significance since all cracks reported to date, with perhaps one exception, are short in length and axially oriented in a flaw-tolerant material.
Further, the effects of wastage by borated water of a creviced area, such as between CRDM penetration and the reactor head, have been evaluated by means of laboratory testing and similar field experience. The findings indicate that any degradation would occur very slowly and, therefore, an event such as ejection of a CRDM continues to be an unlikely event. However, prudence in defense-in-depth and the general design criteria support a position that the industry should develop and implement a well-conceived and comprehensive inspection, repair, anti a:itigation program.
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. In letters of December 17, 1992, and January 15, 1993, WOG submitted a safety analysis to the NRC justifying its basis for continued operation for i:
Westinghouse plants. The WOG will submit a revision to the reports by June 1
1993. The other PWR Owners Groups have endorsed the findings of the WOG l
reports for their facilities, but expect to submit their own safety analyses because their reactor vessel configuration and fabrication differ sufficiently from the Westinghouse plants. The reports will be ready in June 1993.
The industry has comitted to inspect three units in 1994:
(1) Point Beach Unit 1 in the Spring of 1994; (2)
D.C. Cook Unit 2 in the third quarter of 1994; and (3) Oconee Unit 2 in September 1994.
To support the inspection schedule, the following activities are planned:
(1) development and approval of flaw acceptance criteria, January 1994; (2) development of inspection and repair methods and tooling; and (3) development and evaluation by EPRI of potential mitigation methods.
The staff will send an information paper to the Comission after it issues its safety evaluation on this issue. The staff is scheduled to issue its safety evaluation 60 days following the receipt of the two remaining safety analyses of the consequences of postulated cracking of the control rod drive mechanism (CRDM) penetrations from the PWR Owners Groups at the end of June 1993.
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