EA-96-236; 96-249 - Saint Lucie 1 (Florida Power & Light Co.)
EA 96-236 and EA 96-249
Florida Power & Light Company
ATTN: Mr. T. F. Plunkett
President - Nuclear Division
Post Office Box 14000
Juno Beach, Florida 33408-0420
SUBJECT: NOTICE OF VIOLATION (NRC Special Inspection Report Nos. 50-335 and 50-389/96-12)
Dear Mr. Plunkett:
This refers to the inspection completed on July 12, 1996, at your St. Lucie facility. The inspection included a review of selected aspects of your configuration management and 10 CFR 50.59 safety evaluation programs. The results of our inspection were sent to you by letter dated July 26, 1996. A closed, predecisional enforcement conference was conducted in the Region II office on August 19, 1996, with you and members of your staff to discuss the apparent violations, the root causes, and your corrective actions to preclude recurrence. A letter summarizing the conference was sent to you by letter dated September 11, 1996.
Based on the information developed during the inspection and the information you provided during the conference, the NRC has determined that violations of NRC requirements occurred. The violations are cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding them are described in detail in the subject inspection report.
The violation in Part I of the Notice involves your failure to recognize an unreviewed safety question related to the implementation of a valve lineup change to the Emergency Diesel Generator (EDG) fuel oil transfer system. Specifically, in July 1995, the licensee implemented a change to the 2B EDG system to permit closing of a manual isolation valve from the Diesel Fuel Oil Storage Tank to the day tanks in order to minimize fuel oil ground leakage between the two tanks. As part of the change, the licensee instituted administrative measures including dedication of a non-licensed operator and procedural revisions to assure timely opening of the valve following an EDG start. Although a safety evaluation performed to evaluate this change concluded that the probability of loss of the 2B3 emergency bus increased by six percent, it erroneously concluded that no increase in the probability of a component failure was created. In addition, the NRC has concluded that two new failure modes were introduced by the change: (1) potential failure of the operator to unisolate the fuel oil line and (2) failure of the manual isolation valve to open. Therefore, both the possibility for a malfunction of a type different than any evaluated previously in the Updated Final Safety Analysis Report (UFSAR) was introduced, and the probability of a failure of a component important to safety was increased, representing a valid unreviewed safety question.
At the conference, you stated that a safety evaluation was prepared for this change consistent with Florida Power and Light Company procedures and industry guidance (NSAC-125). However, NRC's position with respect to an "increase in probability" differs. Although the NRC recognizes in this case that the increase in probability of component failure was small, a normally passive component was made active and an absolute increase in probability was realized. Notwithstanding the small probability increase, the violation in Part I of the Notice is of significant regulatory concern because a change was made to the EDG system resulting in the emergence of an unreviewed safety question for which a license amendment and NRC approval was not sought. Further, such failures to comply with the requirements of 10 CFR 50.59 resulted in facility operations which depart from the licensing and or design bases described in the UFSAR. Therefore, the violation in Part I of the Notice is classified in accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, as a Severity Level III violation.
In accordance with the Enforcement Policy, a base civil penalty in the amount of $50,000 is considered for a Severity Level III violation. Because your facility has been the subject of escalated enforcement actions within the last two years1, the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process described in Section VI.B.2 of the Enforcement Policy. In this case, the NRC concluded that it is not appropriate to give credit for Identification because the violation was discovered by the NRC. With regard to consideration for Corrective Action, at the conference you stated that your actions related to the violation in Part I of the Notice included revision of engineering safety evaluation guidance to clarify the definition of an increase in probability and issuance of a technical alert to all engineers regarding this issue. Further, although not directly related to this violation, additional emphasis has been placed on the importance of 10 CFR 50.59 and the UFSAR. Your recent actions in this regard include: (1) 10 CFR 50.59 reviewer certification; (2) additional 10 CFR 50.59 training for designated staff; (3) 10 CFR 50.59 procedural enhancements; and (4) implementation of the UFSAR Review Project. Based on the above, the NRC determined that credit was warranted for Corrective Action, resulting in the base civil penalty.
As a result of these considerations, a civil penalty of $50,000 would normally be warranted for this Severity Level III violation. However, in this case, you did perform a 50.59 evaluation and promptly thereafter communicated with the NRC staff and discussed your plans to reposition the fuel oil transfer isolation valve, as well as your preparatory and compensatory measures to minimize the potential for system failure. Accordingly, under the circumstance of this case, a civil penalty is not warranted. I have been authorized, after consultation with the Director, Office of Enforcement, and the Deputy Executive Director for Nuclear Reactor Regulation, Regional Operations and Research, to exercise enforcement discretion, in accordance with the guidance set forth in Section VII.B.6 of the Enforcement Policy, and not propose a civil penalty in this case.
Violations A and B described in Part II of the Notice have been categorized at Severity Level IV. The violations involve four instances where you failed to effectively incorporate design changes into plant operating procedures or drawings. These violations were NRC identified and are of concern because of the potential for misleading operators and the similarity of the violations to annunciator response procedure deficiencies identified during previous inspections. The fifth apparent example of the configuration management violation discussed at the conference involved your failure to incorporate properly the spent fuel pool heat load calculation into operational procedure limitations prior to initiating core off-load. For this issue, the NRC has decided to exercise discretion and characterize the violation as non-cited (NCV 50-335/96-12-01) in accordance with Section VII.B.1 of the Enforcement Policy. Specifically, you identified the violation and promptly instituted appropriate corrective action.
NRC has concluded that no violation occurred with respect to the three additional apparent failures to comply with 10 CFR 50.59 addressed in the subject inspection report and discussed at the conference. Specifically, (1) the Unit 2 Control Element Drive Mechanism Control System Enclosure was not required to be included in the UFSAR, and installation and subsequent modifications did not require 10 CFR 50.59 safety evaluations; (2) the configuration of a temporary fire pump placed in stand-by during the 1996 Unit 1 refueling outage did not require a 10 CFR 50.59 evaluation in that the configuration was as described in the UFSAR (i.e., the discharge valve was closed and the pump was isolated from the system); and (3) the failure to perform a 10 CFR 50.59 safety evaluation to change the setpoints and procedures for operating the fuel hoist was identified and corrected by you prior to actual fuel movement. This letter closes any further NRC action on these matters.
You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response will be placed in the NRC Public Document Room (PDR). To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction.
Sincerely, Original Signed by Stewart D. Ebneter Stewart D. Ebneter Regional Administrator
Docket Nos. 50-335, 50-389
License Nos. DPR-67, NPF-16
Enclosure: Notice of Violations
J. A. Stall
Site Vice President
St. Lucie Nuclear Plant
P. O. Box 128
Ft. Pierce, FL 34954-0128
H. N. Paduano, Manager
Licensing and Special Programs
Florida Power and Light Company
P. O. Box 14000
Juno Beach, FL 33408-0420
Plant General Manager
St. Lucie Nuclear Plant
P. O. Box 128
Ft. Pierce, FL 34954-0128
E. J. Weinkam
Plant Licensing Manager
St. Lucie Nuclear Plant
P. O. Box 128
Ft. Pierce, FL 34954-0218
J. R. Newman, Esq.
Morgan, Lewis & Bockius
1800 M Street, NW
Washington, D. C. 20036
John T. Butler, Esq.
Steel, Hector and Davis
4000 Southeast Financial Center
Miami, FL 33131-2398
Office of Radiation Control
Department of Health and
1317 Winewood Boulevard
Tallahassee, FL 32399-0700
Jack Shreve, Public Counsel
Office of the Public Counsel
c/o The Florida Legislature
111 West Madison Avenue, Room 812
Tallahassee, FL 32399-1400
Joe Myers, Director
Division of Emergency Preparedness
Department of Community Affairs
2740 Centerview Drive
Tallahassee, FL 32399-2100
Thomas R. L. Kindred
St. Lucie County
2300 Virginia Avenue
Ft. Pierce, FL 34982
Charles B. Brinkman
Washington Nuclear Operations
ABB Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 3300
Florida Power and Light Company Docket Nos. 50-335, 50-389 St. Lucie Nuclear Plant License Nos. DPR-67, NPF-16 EA 96-236 and 96-249
During an NRC inspection completed on July 12, 1996, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions," NUREG-1600, the violations are listed below:
I. 10 CFR 50.59, "Changes, Tests and Experiments," provides, in part, that the licensee may make changes in the facility as described in the Safety Analysis Report (SAR) without prior Commission approval, unless the proposed change involves an unreviewed safety question. A proposed change shall be deemed to involve an unreviewed safety question if the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR may be increased, if a possibility for an accident or malfunction of a different type than any evaluated previously in the SAR may be created, or if the margin of safety as defined in the basis for any technical specification is reduced.
Contrary to the above, in July 1995, the licensee made a change to the facility which involved an unreviewed safety question without prior Commission approval. Specifically, the 2B Emergency Diesel Generator (EDG) fuel oil line was manually isolated to secure a through-wall fuel oil leak. In taking this action, the licensee introduced two new failure modes for the 2B EDG, which both increased the probability of occurrence of a malfunction of the EDG above that previously evaluated in the SAR and the possibility for malfunction of a different type than any evaluated previously in the SAR, resulting in an unreviewed safety question. (01013)
This is a Severity Level III violation (Supplement I)
II. 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Criterion III requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis for safety-related structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions.
Florida Power and Light Company Topical Quality Assurance Report, TQR 3.0, Revision 11 implements these requirements. Section 3.2, "Design Change Control," provides, in part, that design changes shall be reviewed to ensure their implementation is in each case, coordinated with any necessary changes to operating procedures. In addition, Section 3.2.4, "Design Verification," provides, in part, that design control measures shall be established to verify the design inputs, design process, and that the design inputs are correctly incorporated into the design output.A. Contrary to the above, the licensee failed to coordinate design changes with the necessary changes to operating procedures as evidenced by the following examples:
- Plant Change/Modification (PC/M) 109-294, "Setpoint Change to the Hydrazine Low Level Alarm (LIS-07-9)," was completed on January 6, 1995, without ensuring that affected Procedure ONOP 2-0030121, "Plant Annunciator Summary," was revised. This resulted in Annunciator S-10, "HYDRAZINE TK LEVEL LO," showing an incorrect setpoint of 35.5 inches in the procedure.
- PC/M 268-292, "Intake Cooling Water Lube Water Piping Removal and Circulatory Water Lube Water Piping Renovation," was completed on February 14, 1994, without ensuring that affected Procedure ONOP 2-0020131, "Plant Annunciator Summary," was revised. This resulted in the instructions for Annunciator E-16, "CIRC WTR PP LUBE SPLY BACKUP IN SERVICE," incorrectly requiring operators to verify the position of valves MV 21-4A and 4B following a safety injection actuation system signal to ensure they were de-energized and had no control room position indication.
- PC/M 275-290, "Flow Indicator/Switch Low Flow Alarm and Manual Annunciator Deletions," was completed on October 28, 1992, without ensuring that affected Procedure ONOP 2-0030131, "Plant Annunciator Summary," was revised. This resulted in the instructions for safety-related Annunciators LA-12, "ATM STM DUMP MV-08-18A/18B OVERLOAD/SS ISOL," and LB-12, "ATM STM DMP MV-08-19A/19B OVERLOAD/SS ISOL," incorrectly requiring operators to check Auto/Manual switch or switches for the manual position. (02014)
This is a Severity Level IV violation (Supplement I).
B. Contrary to the above, the licensee failed to assure that the design of the Circulating and Intake Cooling Water System was correctly translated into plant drawings. Specifically, during implementation of PC/M 341-192, "Intake Cooling Water Lube Water Piping Removal and Circulatory Water Lube Water Piping Renovation," the as-built Drawing No. JPN-241-192-008 was not incorporated into Drawing No. 8770-G-082, "Flow Diagram Circulating and Intake Cooling Water System," Revision 11, Sheet 2, issued May 9, 1995, for PC/M 341-192. This resulted in Drawing No. 8770-G-082 erroneously showing valves 1-FCV-21-3A and 3B and associated piping as still installed. (03014)
This is a Severity Level IV violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201, Florida Power & Light Company is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region II, 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.
Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.
Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. 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.
Dated at Atlanta, Georgia
this 19th day of September 1996
1. A Severity Level III problem and proposed civil penalty of $50,000 were issued on March 28, 1996 (EA 96-040) related to a reactor coolant system boron dilution event. A Severity Level III violation and proposed civil penalty were issued on November 13, 1995 (EA 95-180) related to inoperable power operated relief valves.