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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.
- 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)
- 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)
- 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)
- 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)
- 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)
- 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:
- 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.
- 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.
- 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.
- 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.
- 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:
- 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.
- 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.
- 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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