United States Nuclear Regulatory Commission - Protecting People and the Environment

EA-96-365 - Crystal River 3 (Florida Power Corp.)

March 12, 1997

EA 96-365; EA 96-465; & EA 96-527

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 Nos. 50-302/96-12 AND 50-302/96-19)

Dear Mr. Anderson:

This refers to NRC inspections completed on December 6, 1996, at Florida Power Corporation's (FPC) Crystal River Unit 3 nuclear facility. During these reviews, the NRC examined a number of issues related to your implementation of the Engineering program at Crystal River. The results of the inspections were formally transmitted to you by letters dated November 4, 1996 and January 7, 1997. An open predecisional enforcement conference was conducted in the Region II office on January 24, 1997, with members of your staff to discuss the apparent violations, the root causes, and corrective actions to preclude recurrence. A summary of the conference was sent to FPC by letter dated January 31, 1997.

Based on the information developed during the inspections and the information you provided during the conference, the NRC has determined that violations of NRC requirements occurred. The violations are discussed in detail in Enclosure 1. Enclosure 2 provides the Notice of Violation (Notice). The circumstances surrounding the violations are described in the subject inspection reports.

The violations identified during our inspections indicate a broad spectrum of problems existed in FPC's Engineering program. Our review of these violations disclosed major weaknesses in three specific areas. Six violations were identified for failure to meet the requirements of 10 CFR 50.59. These violations have been categorized in the aggregate as a Severity Level II problem, identified as Violation A in the enclosed Notice, due to the regulatory and safety significance associated with the programmatic deficiencies in the performance of 10 CFR 50.59 safety evaluations and operation of the Crystal River facility with a number of unidentified unreviewed safety questions (USQ) that impacted the design margin and operability of safety-related equipment. Violation B indicates programmatic weaknesses with the implementation of measures to ensure that regulatory and plant design basis requirements are met. Violation B has been categorized as a Severity Level III violation. Violation C involves three examples of untimely and inadequate corrective actions resulting in the failure to identify significant USQs and containment integrity issues and has been categorized as a Severity Level III violation.

FPC's significantly poor performance in these areas is evidence of a systemic breakdown in control of engineering processes at the Crystal River facility. Violation A is of particular concern because of the regulatory significance associated with the failure to implement an adequate 10 CFR 50.59 program and the potential safety consequences that could result from the introduction of a number of USQs that significantly reduced the design margin of certain safety-related systems. Failure to resolve these USQs is a serious safety concern because, in certain accident scenarios, vital equipment could have failed resulting in the loss of the one train of onsite emergency power and unavailability of an emergency feedwater pump. Reactor safety also could have been compromised due to inadequate boron precipitation control in the event of a loss of coolant accident. In addition, failure to identify many of these USQs indicates that your staff failed to recognize design conditions that seriously degraded the margin of safety in several safety-related systems. With regard to Violation B, we are concerned that numerous examples were identified where FPC engineering design control programs lacked sufficient rigor to ensure that design inputs and controls properly maintained the design and licensing bases. Similar weaknesses in the engineering program were identified previously in an escalated enforcement action (EA 95-126), dated July 10, 1996. In addition, FPC failed to take advantage of several opportunities to correct the root causes of these violations prior to NRC involvement and therefore, missed the opportunity to resolve the USQs related to the emergency diesel generator and identify and correct containment configuration control deficiencies.

In accordance with the Enforcement Policy, civil penalties would normally be considered for the Severity Level II problem and Severity Level III violations. However, I have been authorized, after consultation with the Director, Office of Enforcement, and the Deputy Executive Director for Regulatory Effectiveness, Program Oversight, Investigations and 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. It should be stressed that the NRC considered the proposed imposition of a significant civil penalty in this case. However, the NRC has concluded that discretion is appropriate in that: (1) NRC issued a $500,000 civil penalty on July 10, 1996 (EA 95-126) which included sanctions for engineering violations; (2) following NRC identification of the current issues, FPC voluntarily extended the shut down of the Crystal River facility and developed a comprehensive program for problem identification and correction; (3) FPC has demonstrated that remedial action will be taken to ensure reestablishment of design margins for plant systems prior to plant restart; and, (4) FPC's decision to restart the Crystal River facility requires NRC concurrence in accordance with a Confirmatory Action Letter issued on March 4, 1997. FPC's corrective actions will include: (1) completion of a comprehensive restructuring of management; (2) completion of in-depth reviews and corrective actions to ensure compliance with the design bases of the facility; and, (3) implementation of broad and in-depth engineering program changes.

The exercise of this discretion is intended to recognize that FPC has placed the unit in a safe configuration and has committed significant resources to identify and correct deficient conditions. Nonetheless, the NRC must emphasize that failure to implement your improvement plans successfully and substantially improve performance in the engineering area at Crystal River could lead to more significant regulatory sanctions and substantially delay NRC concurrence with restart of Crystal River Unit 3.

As discussed with you at the predecisional enforcement conference, we believe it is imperative that FPC conduct rigorous reviews to ensure that the extent of the deficiencies are bounded and root causes of the conditions are well understood. It is also important to implement long lasting, comprehensive corrective actions that not only correct the deficiencies identified but also strengthen the engineering organization and design review processes. Effective measures to ensure the qualifications of the engineering staff and strong management oversight of the engineering process are also key elements in precluding recurrence of the violations. As part of the responsibility to ensure safe plant operation, the NRC expects licensees with identified programmatic breakdowns to reestablish, without delay, a high level of confidence that design requirements are correctly reflected in the installation and operation of plant equipment and that licensee staff is fully qualified and dedicated to ensuring safe operation of the facility. As discussed at the predecisional enforcement conference, FPC has committed to provide the NRC with additional information as implementation of its improvement plans proceeds to ensure that design margins are appropriately re-established. The NRC will continue to monitor implementation of the improvement program and rigorously review your preparations for restart of Crystal River Unit 3.

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 enclosures, 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

                            Luis A. Reyes
                            Regional Administrator

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

Enclosures:
1. Description of Violations
2. 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

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
B&W Nuclear Technologies
1700 Rockville Pike, Suite 525
Rockville, MD 20852-1631


DESCRIPTION OF VIOLATIONS

A. Inadequate Implementation of 10 CFR 50.59

The NRC determined that a number of unreviewed safety questions (USQs) had been created by plant modifications at Crystal River, and the USQs were not identified in the safety evaluations performed. Each failure to perform an adequate evaluation to determine if a proposed change to the facility or procedures constitutes a USQ is a violation of 10 CFR 50.59. The NRC views each failure to meet the requirements of 10 CFR 50.59 where a USQ existed and the required license amendment was not sought as a significant regulatory issue.

The USQs identified included increases in the emergency diesel generator (EDG) loads for certain accident scenarios in excess of three design load limits which reduced the margin of safety as defined in the Final Safety Analysis Report (FSAR) and Technical Specification (TS) bases; hydraulics and potential net positive suction head (NPSH) problems created by a change to the initiation logic of the alternate steam admission valve for the turbine-driven emergency feedwater (EFW) pump resulting in an increase in the probability of damage to the pump; disabling of one of the automatic steam supplies to the turbine-driven EFW pump, reducing the reliability and increasing the probability of failure of the pump; and, a change to the primary method credited for prevention of boron precipitation in the core following a postulated loss of coolant accident.

The safety consequences of these violations included the potential loss of the "A" train EDG, potential failure of the turbine-driven EFW pump, and potential inadequate control of boron precipitation in the core in the event of a loss of coolant accident. The violations also introduced significant errors in the defined licensing envelope of the plant because documents defining the licensing bases were not updated or were inaccurate.

Commensurate with the regulatory significance of the programmatic deficiencies in the evaluation of plant modifications and procedure changes under the 10 CFR 50.59 program and the potential safety consequences of operating with a number of unidentified USQs, these violations are classified in the aggregate in accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, as a Severity Level II problem.

B. Inadequate Design Control

Five examples of the failure of FPC's engineering design control programs to assure that design inputs and controls properly maintained the design and licensing bases were identified. These examples indicated a significant breakdown in the control of engineering processes at Crystal River. The failure to establish measures to ensure that regulatory requirements and the design basis are correctly translated into specifications, procedures and instructions is a violation of 10 CFR 50, Appendix B, Criterion III, Design Control.

Design control requirements were violated in the following cases: (1) the FSAR, the Enhanced Design Basis Document, and the TS Bases were not updated with regard to operation of the turbine-driven EFW pump (EFP-2) for certain accident scenarios resulting in inadequate evaluation of EDG loading; (2) automatic opening of Valve ASV-204 was disabled preventing the auto-start of EFP-2 in certain accident scenarios; (3) the heat input to the nuclear closed cycle cooling system (SW) heat exchangers was not correctly modeled; (4) unverified calculations were used to support modifications to the EFW system; and, (5) configuration control of containment penetrations was lost in that a significant number of valves and blind flanges were not included in surveillances required to ensure containment integrity. Therefore, this violation has been categorized in accordance with the Enforcement Policy, NUREG-1600, at Severity Level III.

C. Failure to Take Timely and Comprehensive Corrective Actions

Three examples were identified involving the failure to implement timely and appropriate corrective actions and to ensure corrective actions were adequate to preclude recurrence of significant conditions adverse to quality. Failure to take appropriate corrective actions resulted in the creation of USQs and led to reductions in safety margins as discussed below.

The potential for EDG loading in excess of TS limits, identified in May and July of 1996, had not been corrected as of October 11, 1996. Corrective actions identified in a 1994 Licensee Event Report for failure to test containment penetrations in accordance with TS requirements, were inadequate to prevent a recurrence, resulting in failure to identify that numerous additional valves/blind flanges were omitted from surveillance procedures and therefore, were not verified to be closed.

The failure to establish measures to assure that conditions adverse to quality are promptly corrected is a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. Therefore, this violation has been categorized in accordance with the Enforcement Policy, NUREG-1600, at Severity Level III.


NOTICE OF VIOLATION
Florida Power Corporation                          Docket No. 50-302
Crystal River Nuclear Plant                        License No. DPR-72  
Unit 3                                             EA 96-365, 96-465, 96-527

During NRC inspections completed on December 6, 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:

A. 10 CFR 50.59, "Changes, Tests and Experiments," provides, in part, that the licensee may make changes in the facility or procedures as described in the safety analysis report (SAR) without prior Commission approval, unless the proposed change involves a change in the Technical Specifications (TS) or an unreviewed safety question (USQ). A proposed change shall be deemed to involve a USQ 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 TS is reduced. 10 CFR 50.59 further requires that a written safety evaluation be documented providing the bases for a determination that the changes do not involve a USQ.

The TS bases for TS 3.8.1, AC Sources - Operating, states that the service rating of the emergency diesel generator (EDG) is, in part, 3251 to 3500 kilowatts (KW) on a cumulative 30 minute basis.

The Final Safety Analysis Report (FSAR), Rev. 19, dated December 21, 1994, Section 8.2.3, Sources of Auxiliary Power, provides the load ratings for both EDGs, including a 2851 - 3000 KW cumulative 2000 hour rating and a 3251 - 3500 KW cumulative 30 minute rating. (The maximum load rating shown for any period of time is 3500 KW). It also states that the "A" EDG auto-connected load is within the 2000 hour rating at one minute into the scenario. FSAR, Rev. 20, dated April 1, 1994, Table 8-1, Emergency Diesel Generator "A" Auto & Manually Connected Loads, lists the largest auto-connected load as make-up pump 1A (615.5 KW). This FSAR information remained current through 1996.

The FSAR, Rev. 10, dated July 1, 1988, Table 8-1, Emergency Diesel Generator "A" Auto & Manually Connected Loads, lists the largest auto-connected load as make-up pump 1A (615.5 KW). This FSAR information remained current through 1990.

The FSAR, Rev. 10, dated July 1, 1988, Section 10.5, Emergency Feedwater (EFW) System, states that upstream of the turbine-driven emergency EFW pump turbine steam supply line, there are redundant, normally closed direct current (DC) motor operated valves (ASV-5 and ASV-204) which are opened upon actuation from the emergency feedwater initiation and control (EFIC) system. FSAR, Rev. 8, dated July 1, 1987, Section 7.2.4, Emergency Feedwater Initiation and Control, states that the EFIC trip module located in the "A" cabinet actuates the "A" train of EFW (motor-driven pump) and the trip module located in the "B" cabinet actuates the "B" train of EFW (turbine-driven pump)(EFP-2). This FSAR information was the first description of the EFIC system, and it remained current through 1992. Section 7.2.4 of the FSAR, was revised on January 17, 1993, Rev. 18, as follows: "The trip module located in the "A" cabinet starts the "A" train motor-driven EFW pump and the "B" train turbine-driven EFW pump. The trip module located in the "B" cabinet starts only the "B" train turbine-driven EFW pump. The starting of both EFW pumps on "A" train EFIC actuation is necessary to assure that the turbine-driven pump will be operable in the event of a failure of the ES "B" 250/125V DC system coincident with a loss of offsite power and a [engineered safeguards] actuation. Under this scenario, EFP-2 will be relied upon to share the emergency feedwater load with the motor driven emergency feedwater pump in order to decrease the electrical load on diesel generator EDG-3A." This FSAR information remained current through 1996.

  1. Contrary to the above, in April 1996, the licensee made a change to the facility as described in the FSAR, which involved three USQs, without prior Commission approval. Specifically, the modification, installed by Modification Approval Record (MAR) 96-04-12-01 changed the EFW initiation logic to allow the motor-driven EFW pump to provide all EFW during certain analyzed accidents which increased the calculated post-accident motor-driven EFW pump load from about 616 KW to about 666 KW. As a result, the "A" EDG accident loads were in excess of the limits specified in FSAR Section 8.2.3, TS 3.8.1 Basis (3500 KW limit), TS surveillance requirement (SR) 3.8.1.11 Basis (3100 KW one-minute load), and TS SR 3.8.1.8 Basis (616 KW largest single post-accident load that could be rejected). This change reduced the margin of safety as defined in the FSAR and three TS Bases, resulting in three USQs. The 10 CFR 50.59 safety evaluation for this modification was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQs. (01012)

  2. Contrary to the above, in April 1996, the licensee made a change to a procedure as described in the FSAR, which involved three USQs, without prior Commission approval. Specifically, Emergency Operating Procedure EOP-13, EOP Rules, was changed by Rev. 2 to require operators to take manual control of the motor-driven EFW pump to increase EFW flow under certain conditions, resulting in an increase in EFW pump load from about 666 KW to about 713 KW. As a result, the "A" EDG accident loads were in excess of the limits specified in the FSAR and TS 3.8.1 Basis (3500 KW limit), TS SR 3.8.1.11 Basis (3100 KW one-minute load), and TS SR 3.8.1.8 Basis (616 KW largest single post-accident load that could be rejected). This change reduced the margin of safety as defined in the FSAR and three TS Bases, resulting in three USQs. The 10 CFR 50.59 safety evaluation for this procedure change was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQs. (01022)

  3. Contrary to the above, in June 1990, the licensee made a change to a procedure as described in the FSAR, which involved a USQ, without prior Commission approval. Specifically, Operating Procedure OP-402, Makeup and Purification System, was changed by Rev. 64 to allow operators to select, for Engineered Safeguards, the swing "B" makeup pump to either EDG. This resulted in an increase in the largest single post-accident load on the "A" EDG, from 616 KW ("A" makeup pump) to 691 KW ("B" makeup pump). The 691 KW exceeded the largest single post-accident load, that could be rejected by the "A" EDG, specified in the FSAR (616 KW) and in TS SR 3.8.1.2.2 (515 KW). This change in the largest single post-accident load required a TS change which was not made, and therefore resulted in a USQ. The 10 CFR 50.59 safety evaluation for this procedure change was inadequate in that it did not address electrical loading effects on the "A" EDG and did not recognize the USQ. (01032)

  4. Contrary to the above, in May 1987 and in March 1992, the licensee made changes to the facility as described in the FSAR, which involved a USQ, without prior Commission approval. Specifically, modifications TMAR 87-10-09-01 and MAR 87-10-09-01A changed the EFW system electrical power supply for the turbine-driven EFW pump alternate steam admission valve, ASV-204, from "B" train to "A" train DC power and changed the automatic opening of ASV-204 from a "B" train to an "A" train EFW initiation signal. The change introduced a USQ in that, in certain accident scenarios, a failure of the "B" battery would cause the turbine-driven EFW pump to go to runout with its flow control valves failed fully open which would increase the probability of failure of the turbine-driven EFW pump. If the event were also concurrent with a loss of offsite power, the "B" EDG would not operate (due to failure of the "B" battery). Also, the licensee's design basis relied on the turbine-driven EFW pump to share the EFW flow requirements with the motor-driven EFW pump in order to maintain the "A" EDG within its loading limits. The plant operated in various modes from 1987 through April 1996 with this design. The 10 CFR 50.59 safety evaluations for the TMAR and MAR were inadequate in that they did not address hydraulics, potential net positive suction head (NPSH) problems, a resulting potential increase in the probability of a malfunction of the turbine-driven EFW pump, or consequential effects on the "A" EDG; and did not recognize the USQ. (01042)

  5. Contrary to the above, in May 1996, the licensee made changes to the facility, which involved a USQ, without prior Commission approval. Specifically, the 10 CFR 50.59 safety evaluation for MAR 96-04-12-01 (installed in May 1996) was inadequate in that the safety evaluation did not identify that removal of the automatic open signal from valve ASV-204 increased the probability of occurrence of a malfunction of equipment important to safety and therefore was a USQ. Removal of the automatic open signal from valve ASV-204 disabled one of the two automatic steam supplies to EFP-2, which reduced the reliability and increased the probability of a failure of EFP-2. (01052)

  6. Contrary to the above, in 1994, the FSAR was revised in Rev. 21, dated December 1, 1994, Section 4.3.10.1, Boron Dilution, to add information on the boron precipitation methods following a loss of coolant accident ((LOCA), and the 10 CFR 50.59 evaluation was inadequate in demonstrating that a USQ did not exist. Specifically, after identifying deficiencies in the active methods (decay heat drop line and the pressurizer auxiliary spray line) used for boron precipitation control, the FSAR and Design Basis Documents were inappropriately changed to specify flow through gaps in the reactor vessel internals (a passive method) as the first and preferred method. This departed from the original licensing basis of the plant. Also, flow through reactor vessel internal gaps had been identified as acceptable to the NRC (Letter dated March 9, 1993) only as a backup method and not as the primary method. (01062)

These violations represent a Severity Level II problem (Supplement I).

B. 10 CFR 50, Appendix B, Criterion III, Design Control, requires in part, that measures be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2, Definitions, and as specified in the license application, are correctly translated into specifications, procedures, and instructions. In addition, 10 CFR 50, Appendix B, Criterion III, requires that design control measures provide for verifying or checking the adequacy of design, by individuals other than those who performed the original design. It also requires that design control measures shall be applied to items such as the following: reactor physics, stress, thermal, hydraulic, and accident analyses.

The licensee's Quality Program commitments, as described in Table 1-3 of the FSAR, states that in all cases, the design verification shall be completed prior to relying on the component, system, or structure to perform its safety-related function.

Contrary to the above, measures were not established to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, procedures and instructions in the following examples:

  1. The design basis information from calculation E91-0026, approved by the licensee's engineering group on May 11, 1989, was not adequately translated into design documents, in that, the FSAR, Enhanced Design Basis Document, and the TS Bases were not updated to state that the turbine-driven EFW pump (EFP-2) was assumed to be running when the motor-driven EFW pump (EFP-1) tripped automatically at 500 psig reactor coolant system pressure.

  2. Design basis information was not correctly translated into the design input requirements for MAR 96-04-12-01, "ASV-204 EFIC Auto Open Removal," in that the previous credit being taken for EFP-2 operating after EFP-1 automatically tripped when RCS pressure decreased to 500 psig during a LOCA concurrent with a loss of offsite power (LOOP) and failure of the train B vital battery was not recognized in the preparation of the MAR. As a result, MAR 96-04-12-01, which was installed in May 1996, removed the train "A" EFW initiation and control (EFIC) automatic open signal from valve ASV-204, one of the two steam supplies to EFP-2, which would have prevented EFP-2 from automatically starting during certain accident scenarios. The design basis was not met from May 1996 through September 1996. In an event, there would have been no EFW for the period of time between the 500 psig actuation signal and when RCS pressure is reduced below the low pressure injection pump shutoff head (approximately 185 psig), when EFW is no longer required for residual heat removal.

  3. On December 6, 1994, the design basis was not correctly translated into procedures in that, Calculation M94-0056, performed to generate Procedure OP-103B, Curve 15, Nuclear Closed Cycle Cooling System (SW) Heat Exchanger Fouling versus Ultimate Heat Sink (UHS) Temperature, did not correctly model the heat input to the SW Heat Exchangers from the Reactor Building Fan Coolers. As a result, Curve 15 allowed a larger number of SW Heat Exchanger tubes to be blocked, which could have resulted in the SW Heat Exchanger outlet temperature exceeding the 110o F limit during accident conditions and the system not being capable of removing design basis heat from safety-related equipment.

  4. Regulatory requirements were not translated into procedures and the licensee failed to provide measures to verify the adequacy of design by an individual other than those who performed the original design. Specifically, Engineering Procedure NEP-210, Modification Approval Records, Rev. 15, dated January 16, 1996, was inadequate in that it allowed unverified calculations to be relied upon to support modification installation and return to service. As a result, REA 96-047, EDG Loading Case Study, was not verified and was used to support modification MAR 96-04-12-01 approval in April 1996 which contributed to the introduction of three USQs related to EDG loading.

  5. As of June 5, 1996, design basis information was not correctly translated into TS Surveillance Procedures (SP) SP-324, Containment Inspection, SP-341, Monthly Containment Isolation Valve Operability Check, and SP-346, Containment Penetrations Weekly Check During Refueling Operations. Engineering Procedure NEP-210, Modification Approval Records was inadequate in that it did not provide sufficient guidance to incorporate containment isolation valve surveillance requirements in the review of modifications and calculations. In 1988, 1990, and 1996, modifications were installed that would have required revisions to SP-324, 346, and 341, to include certain valves/blind flanges. In 1991, a reanalysis of two containment penetrations resulted in reclassification of the penetrations such that a revision to SP-341, to include valves/blind flanges in the procedure was necessary. In 1996, a review of surveillance compliance to TSs regarding containment integrity was conducted. This review failed to consider the surveillance requirements for containment penetrations in the context of maintenance conditions in SP-346.

As a result of the modifications, reanalysis and review of surveillance compliance, SP-341 did not include 18 valves/blind flanges in the monthly performance check; SP-324 did not include nine valves/blind flanges in the mode 4 to mode 5 surveillance requirement; and SP-346 did not include 55 valves/blind flanges in the surveillance requirement. (02013)

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

C. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that conditions adverse to quality, such as nonconformances, 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 the corrective action taken to preclude repetition.

Contrary to the above, the licensee failed to correct conditions adverse to quality and failed to take measures to assure that corrective actions were taken to preclude repetition of significant conditions adverse to quality as follows:

  1. Precursor Card 96-2750, dated May 31, 1996, and Problem Report 96-0210, dated July 3, 1996, identified that changes made to the facility in April 1996 introduced EDG loads that were in excess of the TS limits, a significant condition adverse to quality; however, adequate corrective actions were not implemented. The adverse conditions were not corrected as of October 11, 1996. As a result, the plant operated for several months with USQs related to EDG loading.

  2. Problem Report 94-0218, dated June 24, 1994, described a problem where engineers failed to address EDG loading effects of several modifications in the 10 CFR 50.59 evaluations; however, the licensee failed to take adequate corrective actions for this significant condition adverse to quality. As a result, in April 1996, the 10 CFR 50.59 evaluation for modification MAR 96-04-12-01 did not address EDG loading effects and MAR 96-14-12-01 was inappropriately installed and placed in operation with USQs.

  3. On October 12, 1994, the licensee identified that penetrations were not being tested in accordance with TS 3.6.3.3, as reported in Licensee Event Report 94-007; however, the corrective actions taken for LER 94-007, dated November 10, 1994, were not adequate to prevent recurrence, resulting in numerous additional valves/blind flanges that were omitted from the surveillance procedures being identified in 1996. (03013)

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 12th day of March 1997

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