IR 05000395/2005007

Final Significance Determination for a Green Finding (NRC IR 05000395/2005007 Virgil C. Summer Nuclear Station)

ML050700044
Person / Time
Site:	Summer
Issue date:	03/10/2005
From:	Ogle C NRC/RGN-II/DRS/EB
To:	Archie J South Carolina Electric & Gas Co
References
EA-05-008, IR-05-007
Download: ML050700044 (51)

Text

rch 10, 2005

SUBJECT:

FINAL SIGNIFICANCE DETERMINATION FOR A GREEN FINDING (NRC INSPECTION REPORT 05000395/2005007, VIRGIL C. SUMMER NUCLEAR STATION)

Dear Mr. Archie:

The purpose of this letter is to provide you with the Nuclear Regulatory Commissions (NRCs)

final significance determination for an issue at South Carolina Electric and Gas Companys (SCE&G) V. C. Summer Nuclear Station. The issue involved inadequacies in your corrective actions associated with a deficiency in the design of the emergency feedwater (EFW) system flow control valves.

The finding resulted from an assessment of Unresolved Item (URI)0500395/2004009-01 of NRC Inspection Report 05000395/2004009, issued on December 22, 2004. The finding was reviewed further in NRC Inspection Report 05000395/2005006, dated January 14, 2005, and was assessed under the significance determination process as a preliminary White finding (i.e.,

an issue of low to moderate safety significance, which may require additional NRC inspection).

The cover letter to the inspection report informed SCE&G of the NRCs preliminary conclusion, provided SCE&G an opportunity to request a regulatory conference on this matter, and forwarded the details of the NRCs preliminary results for this finding.

At SCE&Gs request, a regulatory conference was conducted with you and members of your staff on February 17, 2005, to discuss SCE&Gs position on this issue. The enclosures to this letter include the list of attendees at the regulatory conference, and copies of the material presented by the NRC and your staff at the conference. In support of the conference, SCE&G also provided a written response dated February 9, 2005.

During the conference and as discussed in your February 9, 2005, response, SCE&Gs presentation focused on the likelihood of the unavailability of the Condensate Storage Tank (CST) due to an F2 tornado, and the probability of a random CST failure. SCE&Gs analysis concluded that an F2 tornado would not render the CST unavailable, and based on plant-specific information, a random tank failure event is not risk significant. As such, SCE&G concluded that the change in Core Damage Frequency (CDF) supports a Green finding. Other factors presented by SCE&G at the conference that contributed to its view that the finding is not risk significant included: its differing view that the reactor will not always trip upon a loss of the CST (as postulated by the NRC); operator training to only introduce service water (SW) into the

SCE&G 2 steam generators as a last resort; very little tubercle material will be released by normal EFW flow rates; and any dislodged material will be pulverized in the EFW pumps. Consequently, the EFW flow control valves will not become plugged as assumed by the NRC.

In addition, by letter dated February 23, 2005, and in combination with the information provided in its February 9, 2005, letter and the material presented at the conference, SCE&G provided its conclusions that an adequate basis exists to allow for the NRC to reconsider the issuance of both the 10 CFR 50, Appendix B, Criteria III and XVI violations. In summary, SCE&G stated that the EFW flow control valve specification, selection, and procurement was originally prepared in accordance with prudent and generally accepted design practices, based on the known conditions of the process fluid, including SCE&G operating experience. Therefore, the subject design was developed in accordance and consistent with regulatory requirements. In addition, SCE&G concluded that corrective actions that have been considered and implemented have been appropriate and timely. SCE&G also stated that additional corrective actions currently being pursued to address industry and NRC staff concerns are considered to be enhancements and not the result of ineffective corrective actions.

After considering the information developed during the inspection, the information in SCE&Gs written responses, and the information presented at the conference, the NRC has concluded that the final significance of the finding is appropriately characterized as Green, in the mitigating system cornerstone. In this case, the NRC acknowledges that its preliminary estimate of the change in CDF was White. The NRC considered a slight decrease in the random tank failure probability from that which was assumed in the NRCs preliminary estimate. In addition, SCE&Gs view that the reactor would not always trip upon a loss of the CST appears to be plausible in this case. These factors and other less quantifiable factors discussed by SCE&G would result in a decrease in the change in CDF to a degree slightly less than the Green/White threshold. As such, the NRC has concluded that the significance of this finding should be characterized as Green.

You have 10 business days from the date of this letter to appeal the staffs determination of significance for the identified Green finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2.

Notwithstanding the information provided by SCE&G at the conference and in its written responses, the NRC determined that two violations occurred, involving the requirements of 10 CFR 50, Appendix B, Criterion III and Criterion XVI. The NRC concluded that, at the time of the inspection in November 2004, SCE&G had failed to adequately select and review for suitability of application of materials, parts, equipment and processes that are essential to the safety related functions of the EFW system. In this case, the safety related function of the EFW system, as described in the Updated Final Safety Analysis Report (UFSAR), Section 10.4.9, is for the plant to operate indefinitely, if required, without normal feedwater, and for the EFW system to take suction from the SW system for an indefinite period of time. The design of the EFW system is such that, under certain conditions, the EFW flow control valves could become plugged from tubercles or other debris when aligned to the safety related water supply. The second violation involved SCE&Gs failure to correct a condition adverse to quality wherein the EFW flow control valves were not designed to handle relatively unclean SW and, consequently, could become plugged by tubercles or other debris from SW. The corrective action violation occurred during 1986 through 2004.

SCE&G 3 Because of the very low safety significance of the two violations and because the issues were entered into your corrective action program in CER 0-C-04-3416, the NRC is treating the violations as non-cited violations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. If you contest these NCVs, you should provide a response within 30 days of the date of this letter, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC, 20555-0001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC, 20555-0001; and the NRC Resident Inspector at the V. C. Summer Nuclear Plant.

For administrative purposes, this letter is issued as a separate NRC Inspection Report, No. 05000395/2005007, and the above NCVs are identified as NCV 05000395/2005007-01:

EFW Flow Control Valves Are Susceptible to Plugging by Tubercles or Other Debris from Service Water; and NCV 05000395/2005007-02: Inadequate Corrective Actions in Response to Potential EFW Control Valve Plugging. Accordingly, AV 05000395/2005006-01 and -02 are closed.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Should you have any questions regarding this letter, please contact me at 404-562-4605.

Sincerely,

\\RA\\

Charles R. Ogle, Chief Engineering Branch 1 Division of Reactor Safety Docket No.: 50-395 License No.: NPF-12 Enclosures:

1. List of Attendees 2. Material presented by SCE&G 3. Material presented by NRC

SCE&G 4 cc w/encls.:

R. J. White Nuclear Coordinator (Mail Code 802)

S.C. Public Service Authority Virgil C. Summer Nuclear Station Electronic Mail Distribution Kathryn M. Sutton, Esq.

Winston and Strawn Electronic Mail Distribution Henry J. Porter, Director Division of Radioactive Waste Mgmt.

Dept. of Health and Environmental Control Electronic Mail Distribution R. Mike Gandy Division of Radioactive Waste Mgmt.

S.C. Department of Health and Environmental Control Electronic Mail Distribution Thomas D. Gatlin, General Manager Nuclear Plant Operations (Mail Code 303)

South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station Electronic Mail Distribution Ronald B. Clary, Manager Nuclear Licensing (Mail Code 830)

South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station Electronic Mail Distribution

LIST OF REGULATORY CONFERENCE ATTENDEES NUCLEAR REGULATORY COMMISSION:

L. Plisco, Deputy Regional Administrator, Region II (RII)

V. McCree, Director, Division of Reactor Projects (DRP), RII C. Casto, Director, Division of Reactor Safety (DRS), RII L. Wert, Deputy Director, DRP, RII C. Ogle, Chief, Engineering Branch 1, DRS, RII W. Rogers, Senior Reactor Analyst, DRS, RII K. Landis, Chief, Branch 3, DRP, RII C. Evans, Enforcement Officer and Regional Counsel, EICS, RII R. Schin, Senior Reactor Inspector, DRS, RII J. Moorman, Chief, Operator Licensing Branch, DRS, RII D. Starkey, Senior Enforcement Specialist, Office of Enforcement SOUTH CAROLINA ELECTRIC & GAS COMPANY:

J. Archie, Site Vice President D. Gatlin, Plant Manager M. Fowlkes, Engineering General Manager G. Lippard, Operations Manager B. Whorton, Senior Engineer E. Rumfelt, PRA Engineer R. Clary, Licensing Manager T. Estes, PRA Team Leader R. Guerra, Control Room Supervisor T. Poindexter, Legal Representative Enclosure 1

DRAFT APPARENT VIOLATIONS 1. 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established for the selection and review for suitability of application of materials, parts, equipment and processes that are essential to the safety related functions of structures, systems and components. Implicit in this requirement is that the measures result in the selection of materials, parts, equipment and processes which are suitable.

The design basis as described in Updated Final Safety Analysis Report (UFSAR)

Section 10.4.9, Emergency Feedwater System, states that the service water system provides a safety class backup source of emergency feedwater. The UFSAR further states: "The plant can operate indefinitely, if required, without normal feedwater. The emergency feedwater system (EFW) can take suction from the service water (SW) system for an indefinite period of time."

Contrary to the above, the licensees measures did not ensure that the EFW system can take suction from the SW system for an indefinited period of time.

Instead, the original purchase specification (SP-620-044461-000) for the EFW flow control valves identified the process fluid for the valves as cold condensate.

Consequently, the EFW flow control valves were designed to handle clean 'cold condensate' and were not designed to handle comparatively unclean SW. As a result, if the EFW system were to take suction from the SW system, pieces of tubercles and other debris could plug the EFW flow control valves and cause a common mode failure of the EFW system. The design control aspect of this violation occurred prior to plant licensing in 1982 and existed through 2004.

Note: The apparent violations discussed at this regulatory conference are subject to further review and subject to change prior to any resulting enforcement action.

Enclosure 3

2. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that conditions adverse to quality are promptly identified and corrected.

The licensee identified the potential for SW debris to plug the EFW flow control valves in 1986; conducted ISEG review of the issue in 1986 to 1992; reviewed related operating experience reports in 1988, 2003, and 2004; and photographed corrosion tubercles in the SW pipes to EFW in 2003.

Contrary to the above, as of November 2004, the licensee failed to promptly correct a condition adverse to quality. The licensee failed to correct a condition wherein the EFW flow control valves were not designed to handle relatively unclean SW and, consequently, could become plugged by tubercles and other debris from SW.

Note: The apparent violations discussed at this regulatory conference are subject to further review and subject to change prior to any resulting enforcement action.

Enclosure 3