EA-00-179 - Indian Point 2 (Consolidated Edison Company of New York)

November 20, 2000


Mr. John Groth
Senior Vice President - Nuclear Operations
Consolidated Edison Company of
New York, Inc.
Indian Point 2 Station
Broadway and Bleakley Avenue
Buchanan, NY 10511


Dear Mr. Groth:

The purpose of this letter is to provide you with the final results of our significance determination of the preliminary Red finding identified in the subject inspection conducted between March 7, 2000 and July 20, 2000. The inspection report was sent to you in a letter dated August 31, 2000. This inspection finding was assessed using the significance determination process and was preliminarily characterized as Red, an issue of high safety significance.

This finding involved deficiencies in the overall direction and execution of the 1997 steam generator (SG) inservice examinations at Indian Point 2. Specifically, Consolidated Edison did not identify and correct a significant condition adverse to quality, namely, the presence of primary water stress corrosion cracking (PWSCC) flaws in steam generator tubes, despite opportunities to do so. As a result, tubes with PWSCC were left in service following your 1997 SG inspection until one of these tubes failed on February 15, 2000, when the reactor was at 100% power. As noted in the subject inspection report, the specific opportunities to recognize degraded tubes included the identification of a PWSCC defect, indications of tube denting, and significant eddy current test signal interference. While there were no public health and safety consequences from the tube failure event itself, leaving the degraded tube in service following your 1997 SG inspections resulted in a significant reduction in safety margin during Operating Cycle 14 based on the increased probability of a steam generator tube rupture event.

Our August 31, 2000, letter also provided you an opportunity to attend a Regulatory Conference. The conference, which was open for public observation and transcribed, was held on September 26, 2000, to further discuss your views on this issue. During the conference, your staff discussed an analysis of the probability of a tube rupture, your assessment of the significance of the issue, and measures to prevent recurrence. Also, you indicated that your risk analysis characterized this issue as a Yellow finding, based on your plant-specific analysis of the degraded condition following the 1997 inspection. As a result of your presentation, the NRC requested additional information to support your contention. That additional information, as well as the transcript of the conference and your presentation, were issued by the NRC on October 24, 2000.

The NRC has evaluated the information developed during the inspection, as well as the information you presented during and subsequent to the conference. Based on that evaluation, although the NRC has lowered its calculation of the risk estimate in this case, the NRC revised risk estimate (Enclosure 2) remained above the threshold for classifying this finding as Red, an issue of high safety significance. The NRC recognizes that there is a wide band of uncertainty involved in such risk calculations and additional extensive review could possibly remove some of those uncertainties. Our risk estimate, which classifies the finding as Red, does include a sensitivity analysis that for certain assumptions shows a range of results at the Yellow/Red threshold. However, as noted in our October 10, 2000 letter, the Indian Point 2 facility has been found to have multiple degraded cornerstones. In response to deficiencies at the Indian Point 2 facility, the staff is following guidance in the NRC Action Matrix, which includes oversight of your performance improvement plan and conduct of a significant team inspection.

You have 10 business days from the date of this letter to appeal the staff's determination of significance for the identified Red finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 3.

The NRC has determined that your failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection is a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, as cited in the enclosed Notice of Violation (Notice). The circumstances surrounding the violation were also described in detail in the subject inspection report. In accordance with the NRC Enforcement Policy, NUREG-1600, the Notice of Violation is considered escalated enforcement action because it is associated with a Red finding.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS).

ADAMS is accessible from the NRC Web site at the Public NRC Library.

    Hubert J. Miller
Regional Administrator
Region I

Docket No. 05000247
License No. DPR-26

1. Notice of Violation
2. NRC Significance Determination Analysis

cc w/encls:
A. Alan Blind, Vice President - Nuclear Power
J. Baumstark, Vice President, Nuclear Power Engineering
J. McCann, Manager, Nuclear Safety and Licensing
B. Brandenburg, Assistant General Counsel
C. Faison, Director, Nuclear Licensing, NYPA
J. Ferrick, Operations Manager
C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law
P. Eddy, Electric Division, Department of Public Service, State of New York
T. Rose, NFSC Secretary
F. William Valentino, President, New York State Energy Research
and Development Authority
J. Spath, Program Director, New York State Energy Research
and Development Authority
County Clerk, West Chester County Legislature
A. Spano, Westchester County Executive
R. Bondi, Putnam County Executive
C. Vanderhoef, Rockland County Executive
J. Rampe, Orange County Executive
T. Judson, Central NY Citizens Awareness Network
M. Elie, Citizens Awareness Network
D. Lochbaum, Nuclear Safety Engineer, Union of Concerned Scientists

Consolidated Edison Company of New York
Indian Point 2 Station
  Docket No. 05000247
License No. DPR-26

During an NRC inspection conducted from March 7 through July 20, 2000, a violation of NRC requirements was identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violation is listed below:

10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.

Contrary to the above, despite opportunities during the 1997 Indian Point 2 refueling outage, Con Edison did not fully identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in four Row 2 steam generator tubes, in the small-radius low-row U-bend apex area. In conducting the 1997 steam generator inservice inspection, Con Edison did not adequately account for conditions that adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically, while performing steam generator eddy current test (ECT) examination, during the 1997 outage:

As a result, a minimum of four tubes (with PWSCC flaws in their small radius U-bends) were left in service following the 1997 inspection, until the failure of one of these tubes occurred on February 15, 2000 while the reactor was at 100% power.

This violation is associated with a Red SDP finding.

Pursuant to the provisions of 10 CFR 2.201, Con Edison is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555 with a copy to the Regional Administrator, Region I, and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction. ADAMS is accessible from the NRC Web site at the Public NRC Library. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days.

Dated this 20th day of November 2000

  • a PWSCC defect was identified for the first time, at the apex of one row 2 tube, signifying the potential for other similar cracks in the low-row tubes. However, Con Edison did not adequately evaluate the susceptibility of low-row tubes to PWSCC and the extent to which this degradation existed.
  • indications of tube denting were identified for the first time in low-row tubes at the upper tube support plate (TSP) when restrictions were encountered as ECT probes were inserted into those tubes. Restrictions in 19 low-row tubes signified increased probability of deformed flow slots (hour-glassing) at the upper TSP. Hour-glassing of the upper TSP increases the stresses at the U-bend apex of tubes. These stresses are a prime precursor for PWSCC. However, Con Edison did not adequately evaluate the potential for hour-glassing based on the indications of the low-row tube denting.
  • significant ECT signal interference (noise) was encountered in the data obtained during the actual ECT of several low-row U-bend tubes. This significant noise level reduced the probability of identifying an existing PWSCC tube defect. However, the 1997 SG inspection program was not adjusted to compensate for the adverse effects of the noise in detecting flaws, particularly when conditions that increased susceptibility to PWSCC existed.

Page Last Reviewed/Updated Wednesday, March 24, 2021