ML17179A718

Confirms Plans to Conduct Enforcement Conference on 930222 in Glen Ellyn,Il to Discuss Findings of Insp Identifing Apparent Violation of NRC Requirements at Plant

ML17179A718
Person / Time
Site:	Dresden
Issue date:	02/09/1993
From:	Greenman E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Delgeorge L COMMONWEALTH EDISON CO.
References
EA-93-019, EA-93-19, NUDOCS 9302180009
Download: ML17179A718 (8)

Text

e.

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION tit 799-ROOSEVELT ROAD GLEN ELLYN, ILUNOIS 60137-5927 Docket Nos.

50-237; 50-249 License Nos. DPR-19; DPR-25 EANo.93-019 Commonwealth Edison Company ATTN:

Mr. L. 0. DelGeorge FEB 9 1993 Vice *President, Nuclear Oversight and Regulatory Services Executive Towers West III 1400 Opus Place, Suite 300 Downers Gro~e; IL 60515

Dear Mr. DelGeorge:

This confirms our plans as discussed between Ms. Patricia Lougheed of this office and Ms. Denise Saccomando of your staff to conduct an enforcement conference at 1:00 p.m. (CST) on Monday, February 22, 1993, at the Region III

• office at 799 Roosevelt Road, Building 4, Glen Ellyn, Illinois.

The purpose of this meeting is to discuss the findings of an inspection performed at the Dresden Nuclear Station, Units 2 and 3, which identifi~d apparent violations of NRC requirements.

An inspection report will be.

provided to you prior to our scheduled meeting; a summary of the apparent violations and concerns is provided in Enclosure 1 for your review.

At the enforcement conference, you should be prepared to provide an oral presentation and a concise written handout addressing the root causes and contributing

_factors for the apparent violations and any corrective actions you have taken

. or planned.

This enforcement conference will be open to public observation in accordance with the.Commission's trial program as discussed in the enclosed Federal_

Register notice (Enclosure 2). The purposes of this conference are to discuss the apparent violations, their causes and safety significance; to provide you the opportunity to point out any errors in our inspection-report; and to provide an opportunity for you to present your proposed corrective actions.

In particular, we expect you to address whether an apparent management breakdown contributed to the occurrence of these potential violations.

In addition, this is.an opportunity for you to provide any information concerning your perspective on 1) the severity of th~ issues, 2) the factors that the NRC considers when it determines the amount of a civil penalty that may be assessed in accordance with Section VI.B.2 of the Enforcement Policy, and 3) the possible basis for exercising discretion in accordance with Section VII of the Enforcement Policy.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter ~nd its enclosures will*be placed in the NRC Public Document Room.

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Commonwealth Edison Company 2

J FEB 9 1993 If yo~ have any questions regarding this ~eeting, please contact Ms. Lougheed at (708) 790-5579L

Enclosures:

(1)

Sumniary of Apparent Violations and Concerns (2)

July 10, 1992, Federal Register Notice cc w/enclosures:

M. Lyster, Site Vice President, C. Schroeder, Station Manager J. Shields, Regulatory Assurance Supervisor t

D. Farrar, Nuclear Regulatory Services Manager

• DCD/DBC (RIDS)

OC/LFDCB Resident Inspectors; Dresden LaSalle, Quad Cities, Clinton R. Hubbard J. Mccaffrey, Chief, Public Utilities Division R. Newmann, Office of Public Counsel, State of Illinois Center Licensing Project Manager, NRR State Liaison Officer H.. J. Miller, RIII T. 0. Martin, RIII J. E. Dyer, NRR E. J. Leeds, NRR.

M. L. Jordan, RI! I c~ D. Pederson, RIII S. Stasek, SRI, Davis Besse Sincerely, V)~cl,._ L Edward G. Greenman, Dfrec o Division of Reac~or Projects J; *Lieberman, Director, Office of Enforcement J. Goldberg, Office of the General Council J. Partlow, Director, NRR

SUMMARY

OF APPARENT VIOLATIONS AND CONCERNS Page 1 of 4 You should be prepared to discuss the apparent violatiotis listed bel6w during the enforc~ment ctinference. Additionally, you should be prepared to discuss whether an apparent management breakdown contributed to the occurrence of these potential violatioDs.

• A. --

10 CFR 50.59, Cha-nges, Tests and Experiments allows a licensee to make changes to the facility as described in the safety.analysis report without prior Commission approval unless the proposed change involves a change in the technical spetifications incorporated in the license or an unreviewed safety question. A proposed change is deemed, in part, to involve an unreviewed safety question if the margin of safety ai defined in the basis for any technical specification is* reduced.

Contrary to the above, changes were made to the facility as described in the safety analysis report which involved an unreviewed safety question and prior Commission approval was not obtained. Specifically:

1.

On April 7 and Dece~ber 1, 1992i changes were made to the containment cooling service water {CCSW) system which (a) increased the long-term containment pressure above eight pounds, (b) reduced the containment heat removal capability from 102 million BTU/hr to 77 million BTU/hr, (c) reduced the mini~um number of required CCSW pumps from two to one, and (d) required containment overpressure to achieve emergency core cooling system

{ECCS) pump net positive suction head {NPSH) for a full complement of containment cooling. These changes resulted in a reduction in the margin of safety as defined in the basis for Technical Specifications 3.5.B and 3.7.A.

2.

On March 26, 1988, a modification was approved to the CCSW system which, when implemented, would result in a reduction in the margin of safety as defined in the basis for Technical Specification 3~5.8. Specifically, the low pressure coolant injection {LPCI) heat exchanger tube replacement modification, as approved for implementation, reduced the hea~ removal capability from 102

• million BTU/hr to about 95 million BTU/hr.

Encl osUre l Page 2 of 4 B.

10 CFR Part 50, Appendix B, Criterion Ill, Design Control, requires, in part, that measures be established to assure that.applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

• Contrary to the above, as of January 7, 1993:

1.

Thedesign basis for the emergency diesel generator (EOG) loading capability was not correctly translated into a specification, drawing, procedure, or-instruction in that calculation 7317-33-19-3, "EOG Loading Under DBA Conditions," Revision 7, did not reflect the actual loads that the EDGs could be subject to during use of the plant emergency operating procedures.

2.

The design basis for the licensee's ECCS available NPSH analysis was not correctly tr~nslated into a specification; drawing; procedure, or instruction in that calculation NED-M-MSD-43 "Dresden LPCI NPSH. EvaluatioA-Post DBA-LOCA;" Revision 0, did not reflect the actual available NPSH that the ECCS pumps would be subject to during use of the emergency operating procedures.

C.

10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, thit a program be established to assure that all testingi required to demonstrate that structures, systems~ and components will perform satisfactorily in se~~ice, is identified and performed in accordance with written test procedures which incorporate the requirements and*

acceptance limits contained in applicable design documents.

Contrary to the above:

1.

From January 1985 _until October 1992, the test control program failed to assure that testing was performed to demonstrate that a

• modification, which allowed the CCSW_ pumps to supply cooling water to the control room ventilation system under accident conditions, would perform adequately in service. The acceptance criteria for the initial test was based exclusively on obtaining the minimum

• required flow to the ventilation system's air handling unit, and failed to verify that the CCSW system's minimum flow rate was not

. affected.

2.

From initial plant startup until April 2, 1992, the test control

.program failed to assure that the CCSW system would perform satisfactorily in service. Specifically, the licensee failed to verify that either the Unit 2 or the Unit 3 CCSW system would meet.

the 7000 gpm train flow requirement at the required 180 psig discharge pressure.

e Enclosure!

Page 3 of 4 D.

10 CFR Part 50, Appendix B, Criteria XVI, Corrective Action, requires, in part, that measures be established to assure that condition~ adverse to qtiality ~nd nonconformances are promptly identified and corrected.

Contrary to the above:

1.

Between April 2, 1992, and January 29, 1993, prompt corrective actions were not taken to rectify a condition adverse to quality.

Specifically, on April 2, 1992, the licensee identified, during post maintenance testing, that the flow rate from the CCSW system was decreased from a required 7000 gpm to 5600 gpm.

However, no action was taken to rest9re the required flow rate.

2.

Between April 6 and May 14, 1992, prompt corrective actions were not taken to rectify a condition adverse to quality involving an incorrect heat exchanger duty which was used f~r the essential containment cooling accident analysis.

E.

10 CFR 50.72(b)(l)(ii) requires, in part, that licensees notify the NRC as soon as p~actical and in all cases within one hour of the oc~urrence of any event or condition during operation that results in the nuclear power plant being:

(A) In an unanalyzed condition that significantly compromises plant safety, (B) In a condition that is outside the design basis tif the plant, or (C) In a condition not covered by the plant's operating or emergency procedures.

10 CFR 50.73(a)(l) requires, in part, that licensees submit a licensee event report (LER) for any event described within paragraph 50.73 within 30 days after the discovery of the event. 10 CFR 50.73(a)(2)(ii)(B) requires, in part, reporting of any event or condition that resulted in a condition outside the design basis of the plant.*

Contrary to the above:

• 1.

On Aprii 2, 1992, operations personnel identified a si~nificant reduction in CCSW flow, a condition outside the design basis of the plant. The licensee did not report the event within one hour of discovery or submit an LER within 30 days.

2.

On May 14, 1992, th~ licensee learned that the LPCI heat exchange~

duty was 9 percent degraded from the value used in the most

• limiting accident analysis, a condition outside the design basis of the plant. The licensee did not report the event within one*

hour of discovery or submit an LER within 30 days.

Page 4 of 4

3.

On December 15, 1992, the licensee learned that neither the decay heat model nor the computer methodology used.in the limiting containment heat removal system accident analysis had been approved by the NRC, a condition outside the design basis of the plant.

The licen~ee did not report the event within one hour of discovery or submit an LER within 30 days.

4.

On January 8, 1993, the licensee learned that a potential 20 minute difference existed between the assumed CCSW starting time, used in the accident analysis, and the CCSW starting time expected from operator training. This was a condition outside the design basis of the plant.

The licensee did not ~eport the event to the NRC within one hour of discovery.

In regard to the last potential violation (Item E), you should also address the apparent lack of effectiveness of corrective actions to a previous violation, several examples of failures to make required notifications (Inspection Report 237/249-92009, EA# 92-088).

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