United States Nuclear Regulatory Commission - Protecting People and the Environment

EA-97-162 - Crystal River 3 (Florida Power Corp.)

June 5, 1997

EA 97-162

Florida Power Corporation
Crystal River Energy Complex
Mr. Roy A. Anderson (SA2A)
Sr. VP, Nuclear Operations
ATTN: Mgr., Nuclear Licensing
15760 West Power Line Street
Crystal River, Florida 34428-6708

SUBJECT:  NOTICE OF VIOLATION AND EXERCISE OF ENFORCEMENT DISCRETION
          (NRC INSPECTION REPORT NO. 50-302/97-06)

Dear Mr. Anderson:

This refers to the inspection conducted March 19-21, 1997, at Florida Power Corporation's (FPC) Crystal River Unit 3 nuclear facility. The purpose of the inspection was to assess changes made to the unit's Final Safety Analysis Report (FSAR) and operating procedures that involved a substantial increase in the number of operator actions necessary to mitigate a design basis small break loss of coolant accident (LOCA). The results of the inspection were formally transmitted to you by letter dated April 17, 1997.

Based on the information developed during the inspection, the NRC has determined that a violation of NRC requirements occurred. The violation is cited in the enclosed Notice of Violation (Notice), and the circumstances surrounding the violation are described in detail in the subject inspection report. The violation involved (1) the failure to identify that the addition of required operator actions to mitigate a design basis small break LOCA constituted an unreviewed safety question (USQ); and (2) the subsequent failure to obtain NRC review and approval of that mitigation strategy. The apparent root causes of the violation were inadequate safety evaluations for procedure and FSAR revisions that added operator actions for design basis small break LOCA mitigation. Specifically, FPC failed to identify that, per 10 CFR 50.59(a)(2), the addition of operator actions to the previously approved design basis accident mitigation strategy may result in an increase in the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR and the possibility of a malfunction of a different type than any evaluated previously in the FSAR. The changes introduced additional opportunities for operator error during small break LOCA mitigation and an increase in the probability of reactor coolant pump seal failure. Accordingly, FPC failed to identify that the changes involved an USQ that required NRC review and approval prior to implementation. Further, when FPC recognized that the additional operator actions were required, it failed to incorporate this latest developed information into an FSAR revision that was intended to reflect this change. This violation is a significant failure to meet the requirements of 10 CFR 50.59, including a failure such that a required license amendment was not sought and, therefore, has been categorized as a Severity Level III violation.

In accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, a civil penalty normally would be considered for a Severity Level III violation. However, I have been authorized, after consultation with the Director, Office of Enforcement, to exercise enforcement discretion in accordance with Section VII.B.(6) of the Enforcement Policy and not propose a civil penalty in this case. The NRC has concluded that discretion is appropriate in that: (1) the Crystal River facility is shutdown for performance reasons including engineering violations such as the ones in this case and those issued on March 12, 1997 (EAs 96-365, 96-465 and 96-527) which involved a Severity Level II problem for the failure to perform adequate reviews pursuant to 10 CFR 50.59; (2) the Crystal River facility will remain shut down until completion of a comprehensive program of improvements in the engineering area; (3) FPC has demonstrated that remedial action is being taken to ensure reestablishment of design margins for plant systems prior to plant restart; (4) NRC issued a $500,000 civil penalty on July 10, 1996 (EA 95-126) which included sanctions for engineering violations; and, (5) FPC's decision to restart the Crystal River facility requires NRC concurrence in accordance with a Confirmatory Action Letter issued on March 4, 1997.

During the NRC review of the safety evaluation for FSAR Revision 23, it was determined that errors in the evaluation are similar to those identified in the March 12, 1997 escalated enforcement action described above. The FSAR Revision 23 safety evaluation contains inappropriate reasoning for determining if a USQ exists. For example, the evaluation states that no change was made to the function of any safety-related equipment when in fact automatic functions assumed in the safety analysis were replaced by manual actions. In addition, the evaluation utilizes industry guidelines that have not been accepted by the NRC. These and other statements are of concern to the NRC because FPC had indicated that this safety evaluation was a final product of its reanalysis effort at the time of the March 1997 inspection. Therefore, FPC is requested to include in its response those corrective actions necessary to ensure that safety evaluations contain sound reasoning and complete analysis to support a determination that a USQ does not exist. The NRC also noted that the FSAR Revision 23 safety evaluation contained a reevaluation of a letdown line break which concluded that the potential offsite dose was greater than stated in the FSAR. The NRC will conduct further reviews of this FSAR change.

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, its enclosure, and your response will be placed in the NRC Public Document Room.

Should you have any questions concerning this letter, please contact us.

                             Sincerely, 

                             Original signed by

                             Luis A. Reyes
                             Regional Administrator

Docket No. 50-302
License No. DPR-72

Enclosure: Notice of Violation
cc w/encls:

John P. Cowan, Vice President
Nuclear Production (SA2C)
Florida Power Corporation
Crystal River Energy Complex
15760 West Power Line Street
Crystal River, FL 34428-6708

B. J. Hickle, Director
Nuclear Plant Operations (NA2C)
Florida Power Corporation
Crystal River Energy Complex
15760 West Power Line Street
Crystal River, FL 34428-6708

David F. Kunsemiller, Director (SA2A)
Nuclear Operations Site Support
Florida Power Corporation
Crystal River Energy Complex
15760 West Power Line Street
Crystal River, FL 34428-6708

R. Alexander Glenn
Corporate Counsel
Florida Power Corporation
MAC - A5A
P. O. Box 14042
St. Petersburg, FL 33733-4042

Attorney General
Department of Legal Affairs
The Capitol
Tallahassee, FL 32304

Bill Passetti
Office of Radiation Control
Department of Health and
Rehabilitative Services
1317 Winewood Boulevard
Tallahassee, FL 32399-0700

Joe Myers, Director
Division of Emergency Preparedness
Department of Community Affairs
2740 Centerview Drive
Tallahassee, FL 32399-2100

Chairman
Board of County Commissioners
Citrus County
110 N. Apopka Avenue
Inverness, FL 34450-4245

Robert B. Borsum
Framatome Technologies
1700 Rockville Pike, Suite 525
Rockville, MD 20852-1631


NOTICE OF VIOLATION
Florida Power Corporation                             Docket No.  50-302
Crystal River Nuclear Plant                           License No.  DPR-72
Unit 3                                                EA 97-162

During NRC inspections conducted March 19-21, 1997, a violation of NRC requirements was identified. In accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions," NUREG-1600, the violation is listed below:

10 CFR 50.59(a)(1) states, in part, that licensees may make changes to the facility or procedures as described in the safety analysis report, without prior Commission approval, unless the proposed change involves an unreviewed safety question (USQ). 10 CFR 50.59(a)(2) states, in part, that a proposed change shall be deemed to involve a USQ (i) if the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased or (ii) if a possibility for a malfunction of a different type than any evaluated previously in the safety analysis report may be created. 10 CFR 50.59 (b)(1), in part, states that the licensee shall maintain records of changes in the facility and of changes in procedures made pursuant to this section. These records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USQ. 10 CFR 50.59(c) states that a licensee who desires to make a change in the facility or procedures described in the safety analysis report which involves a USQ shall submit an application for amendment of his license pursuant to 10 CFR 50.90.

Prior to May 2, 1996, the facility Final Safety Analysis Report (FSAR) described the facility's mitigation strategy for a design basis small break Loss of Coolant Accident (LOCA), and that strategy included two operator actions. Those actions were: initiate high pressure injection (HPI) flow through all four injection lines within 20 minutes (per FSAR Table 6-19); and balance flows in the HPI injection lines within 20 minutes (per FSAR Sections 6.1.3.1.2 and 4.2.2.5.7.2). The NRC had previously (in 1979) approved the use of one operator action to mitigate a design basis small break LOCA, i.e., initiate HPI flow through all four injection lines by 10 minutes after the LOCA.

Contrary to the above, on the dates indicated below, the licensee made changes to the facility and procedures described in the FSAR that involved USQs. The changes involved the addition of the operator actions described below to ensure that the design basis requirements for small break LOCA mitigation were met. The FSAR itself was also changed to include some of the operator actions. These changes were made based on inadequate safety evaluations, and as a result, a license amendment was not sought for conditions that involved USQs.

The facility was changed by analysis in Calculation M96-0032, Reevaluation of High Pressure Injection Requirements During Small Break Loss of Coolant Accidents, dated May 2, 1996, such that additional operator actions were required to mitigate the consequences of a design basis small break LOCA. However, the additional operator actions had not been approved by the NRC to be relied upon for mitigation of a design basis small break LOCA. The operator actions added or changed included:

(1) isolate letdown within 10 minutes,
(2) isolate normal makeup within 20 minutes,
(3) isolate reactor coolant pump (RCP) seal injection within 20 minutes,
(4) isolate a broken HPI injection line within 20 minutes, and
(5) control steam generator level above the Emergency Feedwater Initiation and Control automatic setpoint within 20 minutes.

In summary, this change added operator actions [(1), (2), (3), and (5), above] and changed one operator action [(4) above, which replaced the previous operator action to balance flows in the HPI injection lines within 20 minutes] in the facility's mitigation strategy for a design basis small break LOCA.

Procedures described in the FSAR, i.e., the emergency operating procedures, were changed by Short Term Instructions (STI) 95-0061, effective November 8, 1995 to February 8, 1996; STI 96-0068, effective February 8, 1996 to May 6, 1996; and Revision 4 to Emergency Operating Procedure EOP-03, dated May 2, 1996; to add the operator action to isolate RCP seal injection. The remaining operator actions had been in the emergency operating procedures since the late 1970s, but at least four of them had not been relied upon to satisfy the design basis as stated in the FSAR.

The FSAR was changed by Revision 23, titled "FSAR Revision due to HPI Reevaluation," dated November 18, 1996, to incorporate the results of Calculation M96-0032 into the FSAR. This change included the operator actions listed above, with the exception of action (5). The safety evaluation for FSAR Revision 23 was dated April 30, 1996.

The required safety evaluations that supported Revision 23 to the FSAR, STI 95-0061, STI 96-0068, and Emergency Operating Procedure EOP-03, Rev. 4, were inadequate in that they failed to recognize the introduction of USQs. (There was no separate safety evaluation for Calculation M96-0032.) The inadequacies involved a failure to recognize that the increase in the number of operator actions required to mitigate a design basis small break LOCA introduced the possibility of a malfunction of a different type than any evaluated previously in the FSAR, and also increased the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR. The changes introduced additional opportunities for operator errors. The inadequacies also involved a failure to recognize that the addition of the action to isolate RCP seal injection increased the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR in that it increased the probability of seal failure. Therefore, the changes involved USQs.

The safety evaluation for FSAR Revision 23 was also inadequate in that it failed to address, and failed to ensure that the FSAR included, all of the required operator actions for small break LOCA mitigation that were stated in Calculation M96-0032. Operator action (5) was not addressed by the safety evaluation for FSAR Revision 23 and was not included in the FSAR revision. (01013)

This is a Severity Level III violation (Supplement I).

Pursuant to the provisions of 10 CFR 2.201, Florida Power Corporation (Licensee) 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 Crystal River facility, 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 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 previously 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 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 Action, 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 5th day of June 1997

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