IR 05000302/1993004
| ML20035B637 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/19/1993 |
| From: | Freudenberger, Holmesray P, Landis K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20035B631 | List: |
| References | |
| 50-302-93-04, 50-302-93-4, NUDOCS 9304020249 | |
| Download: ML20035B637 (15) | |
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ATLANTA, GEORGI A 30323 4.....,o
Report No.:
50-302/93-04-Licensee:
Florida Power Corporation 3201 34th Street, South St. Petersburg, FL 33733 Docket No.:
50-302 License No.: DRP-72.
Facility Name:
Crystal River 3 Inspection Conducted: January 17 - February 20 Inspector:
O. P. lms 5[/9 97 P. Holmes-Ray, Senior Resident Inspector Date Signed Inspector:
Q. R, lce 3/19f33 R. Freude ger, esigntInspector Dat Si ned Approved by:
M7z4#h 7 /7 b K. Laridis, Section Chief Date Signed Division of Reactor Projects Accompanying Personnel:
A. Long, Project Engineer, RII SUMMARY Scope:
This routine inspection was conducted by two resident inspectors in the areas of plant operations, security, radiological controls, Licensee Event Reports, and licensee action on previous inspection items. Numerous facility tours were conducted and facility operations observed.
Backshifts inspections were conducted on January 18, 29, 30, February 2, 4 and 6.
Results:
In the area of plant operations, the following was identified:
A walkdown of the steam supply piping to the turbine driven emergency feedwater pump found the system properly aligned.
(paragraph 3.b.)
In the area of maintenance / surveillance:
A diving operation was well controlled.
(paragraph 4.a.)
The licensee's actions in response to the change in differential pressure on raw water pump
"3A" and the identification and analysis of the degraded condition of the
"A" spent fuel pool heat exchanger were appropriate.
(paragraph 4.b.)
Outage planning scope control was improved as compared to previous outages.
(paragraph 4.c.)
9304020249 930319 PDR ADOCK 05000302 G
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.The following open items were reviewed:
VIO 50-302/91-14-01 CLOSED Failure to take prompt corrective
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action to control all control i
complex habitability penetrations as i
required by'10 CFR 50, Appendix B,
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Criteria XVI. (paragraph 6.a)
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VIO 50-302/91-24-01 CLOSED Failure to perform channel-l functional tests of anticipatory.
l reactor trips prior to entering Mode 1. (paragraph 6.b)
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VIO 50-302/92-10-01 CLOSED Failure to perform required review t
and approval of WR instructions.
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VIO 50-302/92-12-01 CLOSED Failure to take corrective action to i
prevent recurrence of a posted high l
radiation boundary area door left
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full open and unmonitored.
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URI 50-302/92-16-02 CLOSED Development and implementation of corrective action plan for Problem
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Report 92-0031. (paragraph 6.e)
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URI 50-302/92-18-02 CLOSED Locking of EDG support systems
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cross-connect valves in the closed
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position. (paragraph 6.f)
LER 91-11 CLOSED Discovery of condition-having potential to compromise long term cooling post-LOCA. (paragraph 5.a)
LER 92-05 CLOSED Emergency diesel generator could l
exceed maximum rating for certain i
MVP operational lineups die to not
analyzing the condition of loss of
offsite power coincidental with
engineered safeguards actuation
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during surveillance testing.
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(paragraph 5.b)
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LER 92-12 CLOSED Inadequate procedure results in j
isolation of residual heat removal j
system. (paragraph 5.c)
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REPORT DETAILS.
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Persons Contacted Licensee Employees
- W. Bandhauer, Nuclear Shift Manager i
G. Boldt, Vice President Nuclear Production f
- J. Colby, Manager, Site Nuclear Engineering Services (Acting)
l E. Froats, Manager, Nuclear ~ Compliance
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- G. Halnon, Manager, Nuclear Plant System Engineering B. Hickle, Director, Nuclear Plant Operations
- M. Jacobs, FPC Public Relations
- H. Koon, Nuclear Maintenance Superintendent i
- K. Lancaster, Superintendent Nuclear Maintenance Work Controls
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- G. Longhouser, Nuclear Security Superintendent
- W. Marshall, Manager, Nuclear Plant Operations
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- P. McKee, Director, Quality Programs
- R. McLaughlin, Nuclear Regulatory Specialist l
- B. Moore, Manager, Nuclear Integrated Scheduling
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- J. Roberts, Assistant Nuclear Chemistry & Radiation Protection i
Superintendent
- S. Robinson, Manager, Nuclear Quality Assessments i
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- W. Rossfeld, Manager, Site Nuclear Services
- J. Terry, Supervisor Site Nuclear Engineering Services
- R. Widell, Director, Nuclear Operations Site Support j
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Other licensee employees contacted included office, operations, engineering, maintenance, chemistry / radiation, and corporate personnel.
NRC Resident Inspectors
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- P. Holmes-Ray, Senior Resident Inspector
- R. Freudenberger, Resident Inspector i
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- Attended exit interview Acronyms and initialisms used throughout this report are listed in'the f
last paragraph.
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2.
Plant Status and Activities
The plant continued in power operation (Mode 1) for the duration of this
inspection period.
On January 22, a meeting was conducted at the Region II Office. The purpose of the meeting was a presentation by FPC on Operations
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Enhancements at Crystal River.
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On January 27 & 28, a representative of the United Kingdom Nuclear
Installations Inspectorate was on site for a plant tour and discussions
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related to the Technical Specification Improvement Program.
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3.
Plant Operations (71707, 93702, & 40500)
Throughout the inspection period, facility tours were conducted to observe operations and maintenance activities in progress. The tours-included entries into the protected areas and the radiologically controlled areas of the plant. During these inspections, discussions were held with operators, health physics and instrument and controls technicians, mechanics, security personnel, engineers, supervisors, and plant management. Some operations and maintenance activity observations were conducted during backshifts.
Licensee meetings were attended by the inspector to observe planning and management activities.
The inspections confirmed FPC's compliance with 10 CFR, Technical Specifications, License Conditions, and Administrative Procedures.
a.
Transformer Oil Spill On January 26, while filling a transformer located within the protected area, approximately one gallon of oil was spilled. A portion of the oil was released via a storm drain to the discharge canal.
Since the spill was reported to the Florida Marine Patrol and the United States Coast Guard, the NRC was notified in accordance with 10 CFR 50.72 (EN 24948).
The spill was quickly contained and cleaned up.
Further inspector review was not warranted.
b.
Emergency Feedwater Pump Steam Supply A walkdown of the steam supply piping to the turbine driven emergency feedwater pump was conducted. The purpose of the walkdown was to verify proper alignment of steam traps to prevent condensation in the steam supply line. The inspection was prompted by a recent similar problems in the industry. The system was properly aligned, and no discrepancies were identified.
c.
Radiological Protection Program Radiation protection control activities were observed to verify that these activities were in conformance with the facility policies and procedures, and in compliance with regulatory requirements.
These observations included:
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Entry to and exit from contaminated areas, including step-off pad conditions and disposal of contaminated clothing;
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Area postings and controls;
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Work activity within radiation, high radiation, and contaminated areas;
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RCA exiting practices; and
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Proper wearing of personnel monitoring equipment, protective clothing, and respiratory equipment.
The implementation of radiological controls observed during this inspection period were proper and conservative, h
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Security Control
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of the licensee's physical security program. The performance of various shifts of the security force was observed in the conduct of daily activities to include: protected and vital areas access controls; searching of personnel, packages, and vehicles; badge issuance and l
retrieval; escorting of visitors; patrols; and compensatory posts.
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addition, the inspector observed the operational status of protected j
area lighting, protected and vital areas barrier integrity, and the
security organization interface with operations and maintenance. A i
detailed walkdown of the perimeter fence, gates, and locking mechanisms l
was performed during the week of February 15.
No performance discrepancies were identified by the inspectors.
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Fire Protection
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Fire protection activities, staffing, and equipment were observed to i
verify that fire brigade staffing was appropriate and that fire alarms, l
extinguishing equipment, actuating controls, fire fighting equipment, i
emergency equipment, and fire barriers were operable.
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t Violations or deviations were not identified.
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Maintenance and Surveillance Activities (62703 & 61726)
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Surveillance tests were observed to verify that approved procedures were j
being used; qualified personnel were conducting the tests; tests were adequate to verify equipment operability; calibrated equipment was l
utilized; and TS requirements appropriately implemented.
l The following test was observed and data reviewed:
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SP-110, Reactor Protection System Functional Testing.
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In addition, the inspector observed maintenance activities to verify j
that correct equipment clearances were in effect; work requests and fire l
prevention work permits, as required, were issued and being followed; j
quality control personnel performed inspection activities as required; j
and TS requirements were being followed.
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Maintenance was observed and work packages were reviewed for the following maintenance activities:
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WR 283109, Dive into Raw Water Pit "A" to search for debris and-l check condition of Raw Water Pump;
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WR 297343, Spent Fuel Heat Exchanger tube plugging, tube removal l
and reassembly; i
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"A" Reactor Cavity Cooling Fan (AHF-2A) rebuild following inservice failure; and
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WR 300876, Replacement of EDG Day Tank Level Instruments.
The following items were considered noteworthy; a.
Raw Water Pit Dive On January 25, a dive into the "A" raw water pit was performed. The purpose of the dive was to inspect the pit for excessive silting and/or unexpected Jebris. These conditions were suspected due to a minor decrease (approximately 3%) in the differential pressure developed during routine surveillance testing of raw water pump 3A.
The licensee
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concluded that the pump remained operable based primarily on measured pump flow. Also, the change in differential pressure was insufficient to indicate a substantial deviation from the baseline pump curve.
Since the change was unexplained the licensee chose to perform an inspection of the raw water pit and pump suction bell using a diver. The inspector reviewed procedure CP-139, Diving Operation Work Permit Procedure, attended the pre-dive meeting, verified the dive clearance, and observed the diving operation.
Some silting was identified and debris was removed from the pit; however, no conditions were identified which accounted for the change in differential pressure.
Following the dive, pump differential pressures remained slightly degraded but within specification. The inspector considered the diving operation to be well controlled and the licensee's actions in response to the change in RWP-3A's differential pressure to be appropriate.
b.
Spent Fuel Heat Exchanger WR 297343 was initiated to repair tube leakage in the "A" spent fuel heat exchanger.
Concurrent with plugging of identified leaking-tubes,
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the licensee performed eddy current testing of accessible portions of the "U" shaped tubes.
Eddy current testing and shell side pressure
testing identified a total of 17 tubes that required plugging. There l
are 148 tubes in the heat exchanger. The tube leaks appeared to be
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primarily due to defects in the "U" bend area of the tubes. Based on the number of tubes which demonstrated degradation and the inability to l
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perform eddy current testing in much of the "U" bend areas, the licensee removed two tubes to perform a failure analysis.
The plugging and/or removal of 17 tubes from the heat exchanger reduced
the heat exchanger's capacity by approximately 11.5%.
Problem Report
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93-0026 was generated based on an engineering evaluation which indicated i
that the overall heat removal capacity of the
"A" train of the spent i
fuel coolant system had been reduced such that the design basis l
conditions would not te met. Specifically, the maximum spent fuel pool
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temperature of 160" F would be exceeded with a full reactor core i
discharge to the spent fuel pools 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown at the end of a two-year operating cycle. The heat load under these conditions
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is significantly greater than the normal heat load of the spent fuel l
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pools. The licensee plans to complete the failure analysis, determine specifications for, and procure a replacement tube bundle prior to the next scheduled refueling outage in the spring of 1994. The "B" spent fuel pool heat exchanger has not exhibited tube leakage and has no-plugged tubes. The inspector considered the licensee's actions related to the identification and analysis of the degraded condition of. the "A" spent fuel pool heat exchanger to be appropriate, c.
Outage Planning Outage scope control has been a major concern to FPC during previous outages. After Refueling Outage 8, in 1992, the outage planners were tasked to developed a program to control outage scope. 'The Mid-Cycle 9 Outage length was established by the time required to complete the
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longest major modification.
Identified work was screened based on
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safety and financial restraints and items not to be accomplished were
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deferred. A deferral list was promulgated for departmental. review.
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addition, the deferral list was reviewed bv the PRC to ensure the priority safety items remained in the outage scope. The PRC plans to i
review further deletions from the outage as they occur.
Compared to
previous outages, these actions were viewed as improvements by the
inspector.
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Overall, surveillance and maintenance activities observed and discussed above were performed in a satisfactory manner in accordance with
procedural requirements and met the requirements of the TS.
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Violations or deviations were not identified.
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5.
Review of Licensee Event Reports (92700)
LERs were reviewed for potential generic impact, to detect trends, and to determine whether corrective actions appeared appropriate.
Events
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that were reported immediately were reviewed as they occurred to e
determine if the TS were satisfied.
LERs were also reviewed in
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accordance with the current NRC Enforcement Policy.
a.
(Closed) LER 91-11: Discovery of condition having potential to
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compromise long term cooling post-LOCA.
t CR-3 was in Mode 5, Cold Shutdown, on November 4,1991. At 4:30 p.m.
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I utility personnel determined from analysis that a condition existed i
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vbich appeared to have the potential to affect long term cooling if a LOCA occurred due to a break in one of the reactor coolant system cold legs.
The basis for CR-3's operating license includes the post-LOCA long term
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cooling analyses performed for B&W plants, which indicates that actions to prevent boron precipitation in the core would not be required until approximately forty days following a cold leg break LOCA. More recent analyses indicated boron precipitation could begin much sooner and actions to prevent boron precipitation would be required within twenty l
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four hours.
Procedures for CR-3 already required these actions within twenty four hours.
Since notification of this problem, additional analysis has shown that flow phenomena within the reactor will prevent the boron concentration from increasing to the solubility limit. Therefore, no boron precipita-tion would occur so no procedural or hardware revisions were required.
This LER is closed.
b.
(Closed) LER 92-05:
Emergency diesel generator could exceed maximum rating for certain MVP operational lineups due to not analyzing the condition of loss of offsite power coincidental with engineered safeguards actuation during surveillance testing.
On April 29,1992, CR-3 was operating at 63% of full power. At 12:30 p.m., it was determined by FPC engineering that during surveillance testing of either EDG with the EDG running in parallel with its respective 4160V ES bus, if a loss of offsite power were to occur coincident with an ES actuation, the EDG engine could be overloaded beyond its maximum rating of 3500 KW and may stall.
In this scenario, the overloading of the affected EDG will only occur if the running MUP is not selected for ES response and is being powered from the ES bus associated with the testing.
If this MVP selection is made, during block loading the running EDG would attempt to power two MVPs as well as the other ES loads associated with that bus. Both EDGs were being maintained in normal standby at the time of identification of this condition, therefore no immediate action was required. The event was reported to the NRC via ENS in accordance with 10 CFR 50.72 (EN#23361).
The inspector reviewed the revisions to procedures SP-354A, Monthly Functional Test of the Emergency Diesel Generator EGDG-1A; SP-354B, Monthly Functional Test of the Emergency Diesel Generator EGDG-1B; and OP-707, Operation of the ES Emergency Diesel Generators. Revisions to the procedures included steps to ensure that if the running MUP is powered from the ES bus to be supplied by the EDG, the running MVP is ES selected. This LER is closed.
c.
(Closed) LER 92-12:
Inadequate procedure results in isolation of residual heat removal system.
On June 27, 1992, CR-3 was in dode 5 (Cold Shutdown) with the "A" train of DHR in service. The plant had been shut down since April 30, 1992, for a scheduled refueling outage.
Fuel had been recently reloaded into the reactor vessel.
I&C technicians were recalibrating one of the RCS pressure instrument strings. The recalibration was necessitated by a recently completed plant mcdification. When the buffer amplifier module
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was removed, one of the valves in the RCS dropline closed. This isolated the suction of the operating DHR pump from the RCS. The pump was turned off to protect it. Cooling water flow to the reactor core was interrupted for approximately ten minutes while power to the instrument string was recovered. The valve was reopened and the pump
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was restarted. A report was made to the NRC in accordance with 10 CFR 50.72 (EN 23747).
This event was the subject of violation 50-302/92-16-01. Corrective action associated with the LER will be reviewed as part of the violation followup.
LER 92-12 is closed.
Violations or deviations were not identified.
6.
Licensee Action on Previously Identified Inspection Findings (92702 &
92701)
a.
(Closed) Violation 50-302/91-14-01:
Failure to take prompt corrective action to control all control complex habitability penetrations as required by 10 CFR 50, Appendix B, Criteria XVI.
On January 12, 1990, with the unit in Mode 1, at full power, a door to the control complex was removed for modification in accordance with WR NU 0257685 and MAR 89-07-02-01. A licensee engineer making a walkdown of the control complex ventilation system noticed the removed door and questioned the operability of the system with the door removed. The door was replaced within eight hours. On April 23, 1990, the licensee determined that the door being removed allowed excessive leakage into the control complex and therefore both trains of the Control Complex Emergency Ventilation System had been inoperable.
LER 90-07 was generated to document this event and stated that, "This event was caused by inadequate implementation of the Control Complex Habitability Study...".
In August 1991, an NRC review of the corrective actions stated in the LER was performed. The corrective actions stated in the LER were to post signs on the control complex doors alerting personnel as to the special nature of the doors and to develop a program for control of all control complex habitability envelope penetrations.
Signs were made and
installed on all control complex doors to require contacting the
operations shift supervisor on duty prior to removal or repair of the
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doors. The inspector verified that the signs were in place; however, no program for the control complex habitability penetrations was in place.
Violation 50-302/91-14-04 was issued for failure to take prompt
corrective action to control all control complex habitability
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In the rer.ponse to the notice of violation, the licensee stated that it
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was judgel that interim instructions were adequate. Therefore, a low priority was assigned which delayed the development of a formal, procedure-based program.
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Corrective actions included the revision of procedure CP-137, Fire Barrier Penetration Breach Report, to include breaches of the control l
complex habitability envelop; the development of TS Interpretation 91-02
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which delineates TS actions related to control complex habitability
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envelop breaches; the development of procedures for the determination of unacceptable seals on doors and penetrations; the development of
procedural guidance for identification and preventive actions to be included in WR that have the potential to violate the control complex _
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habitability envelope; and the incorporation of pertinent information into operator requalification training.
The inspector reviewed CP-137, Breach Authorization Program; TS Interpretation 91-02 SP-805A, Annual Inspection of Plant Fire Doors; SP-407, Fire Barrier Penetration Seals; CP-113B, Work Request Evaluation; and training records for licensed operator requalification
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training lesson ROT-5-49. The corrective actions appear to be sufficient to prevent recurrence. This violation is closed.
b.
(Closed) Violation 50-302/91-24-01: Failure to perform channel
functional tests of anticipatory reactor trips prior to entering
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Mode 1.
l On November 24, 1991, the unit began mode ascension from Mode 3 and achieved Mode 1 at 5 p.m. that day. On November 25, 1991, the unit
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tripped from 20% power due to a secondary plant transient that caused loss of both main feedwater pumps. The anticipatory trip from loss of main feedwater pumps tripped the reactor. On November 26, 1991, the i
licensee discovered that the anticipatory trip portion of the required l
RPS surveillances for loss of main feedwater and loss of main turbine
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had not been performed prior to the November 24, 1991, mode changes.
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The event was attributed by the licensee to personcel error on the part of the responsible shop supervisor, for not assuring that the surveillance was performed within the required time frame. The corrective action included procedure changes to improve controls for I
ensuring shut down surveillances are completed on time.
LER 91-15 was
issued to address this event. The corrective actions associated with
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this violation were reviewed in NRC Inspection Report 50-302/92-23,
detail 6.c.
In that review, the inspector noted that the date of last
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surveillance performance was not included on the mode change checklists.
i The inspector concluded that the shift supervisor could not determine from the information on the checklist whether the completed surveillances were still valid or if the time limits had expired.
Therefore the shift supervisor's review would not necessarily catch the
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type of errors which produced this event because the mode change checklist did not provide this information. The inspector discussed
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this problem with the licensee. The licensee acknowledged the problem
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and intended to revise SP-440, Unit Startup Surveillance Plan, to include surveillance scheduling information.
Revision 65 to SP-440 was issued on January 26, 1993. This revision i
included the completion date as well as window expiration dates for the validity of required surveillances. This item is closed.
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(Closed) Violation 50-302/92-10-01: Failure to perform required t
review and approval of WR instructions.
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The violation involved two examples of work instructions that had not been reviewed and approved by the PRC as delineated in TS 6.8.1 and 6.8.2.
Corrective actions performed by the licensee included a revision to procedure AI-400, Description and General Administration of Plant Procedures,.to provide more definitive criteria to implement the department policy for CR-3 organizations. A revision will be made to maintenance program work process procedure CP-113B, Work Request Evaluation, to assure that the threshold for work instruction review and approval is adequately defined and incorporated into process control procedures and guidelines. Also, the cited work instructions were reviewed by the PRC.
The inspector reviewed the above mentioned procedure revisions. The
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revisions, together with the absence of continued findings in this area during subsequent inspections, indicate that the issue was adequately
addressed by the licensee. This item is closed.
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(Closed) Violation 50-302/92-12-01:
Failure to take corrective action to prevent recurrence of a posted high radiation boundary area door left full open and unmonitored.
l The violation involved the repetitive failure to positively control a posted high radiation area boundary door.
Corrective actions performed by the licensee included; a permanent revision to General Employee Training, lesson plan GT-001, Green Badge / Fitness for Duty / Yellow Badge and GT-003, Green Badge / Fitness for Duty / Yellow Badge Requalification, to strengthen individual awareness of
responsibilities concerning high radiation barriers; closing mechanism i
plant was evaluated by the Radiation Protection department; and the
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installation on appropriate high radiation area doors throughout the Radiation Protection department evaluated control of high radiation
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areas to improve posting and locking relevance to actual radiological
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conditions. Although the door in question provided access to an area
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for which locked access was not required, the licensee has chosen to
continue to control the area as a high radiation area due to the-
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potential for rapidly changing radiological conditions in the area
during postulated plant transients.
l These actions were reviewed by the inspector. Together with the absence
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of continued findings during subsequent inspections, the actions were i
considered sufficient to adequately address the issue. This item is
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closed.
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(Closed) Unresolved Item 50-302/92-16-02: Development and
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implementation of corrective action plan for Problem Report 92-
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0031.
This issue involved a concern regarding the settings of the MOV limit i
switches which control high pressure injection flow following an ES
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actuation. The concern was that there were two open limit switches
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associated with each injection valve and the current method for setting the limits did not clearly define which limit was the ES throttled position and which was the full open limit. The ES throttled position was used to balance flows among the injection lines. The full open position was used to prevent back seating of the valves. Since the HPI valves are plug type throttle valves and the openings in the lower portion of valve cage are smaller than the reactor building recirculation sump screens, the full open limit was provided to allow the operators to open the valves fully, uncovering larger openings in the cage, during the recirculation phase of an accident.
System Engineering review concurred that the present method was not adequate to ensure correct as-left limit switch settings. Based on this conclusion, Problem Report 92-0031 was initiated and WRs were generated to measure as-found limit switch settings prior to the performance of preventive maintenance activities scheduled on the MOVs during the refueling outage.
The as-found data was collected on May 26, 1992.
It indicated that both the ES throttled position and the full open limit switches were set at the industry-standard full-open position of 90% to 95% open.
It appeared that both the open limit switches on each of the four valves were set to the full open position in 1987, when the MOV torque switch bypass modification (MAR 87-03-13-02) was performed.
In that configuration, past performance data existed only to demonstrate that MUP-1B lineup was capable of meeting the TS 4.5.2.g. requirements. Data taken during the performance of SP-414, High Pressure Injection Flow Verification Test, in 1986, showed the TS requirements of 500 gpm total flow and a minimum of 350 gpm for any combination of three out of four injection lines, at a minimum RCS pressure of 600 psig was achieved with'
the injection valves in the full open position and MVP-1B running.
No conclusive data existed to show that MVP-1A or MVP-lc could meet the minimum flow of 350 gpm for any combination of three out of four injection lines with the injection valves full open.
A revised version of SP-414 was performed in Mode 3 during startup from the Refuel 8 outage. This test demonstrated that with any MVP in operation and the HPI valves full open, flows balanced within the acceptance criteria of TS 4.5.2.g were achieved. Therefore, the operation of the HPI system was within its design basis with both sets of open limit switches set to the full open position.
The corrective action plan associated with Problem Report 92-0031, which addresses the cause of the improper MOV limit switch settings, was under development at the end of the report period.. This issue was unresolved pending development and implementation of the complete corrective action pl an.
During this inspection, the inspector reviewed the corrective actions which were related primarily to drawing and procedure revisions. These appeared to be sufficient to prevent recurrence of the problem.
Although all actions were not complete, the inspector considered
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This item is l
closed.
f.
(Closed) Unresolved item 50-302/92-18-02:
Locking of EDG support systems cross-connect valves in the closed position.
This issue involved the potential for a comon mode failure of EDG I
support systems. Cross-connect piping exists between the EDG air start i
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receivers, fuel oil storage tanks and day tanks (See FSAR section 8.2.3.1 page 8 - 20). Two manual cross-connect isolation valves (EGV-25 l
and 26, DFV-47 and 48, DFV-56 and 59) between each of these components
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are normally closed. However, a common mode EDG failure could exist if
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these cross-connect valves were inadvertently opened and a fault
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occurred on either EDG component (e.g., tank rupture, etc.) and both EDGs could thus become inoperable. This unresolved item was opened to
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track the resolution of whether the cross-connect valves should be l
locked-closed valves. As documented in NRC IR 50-302/92-18-02, Surveillance procedure SP-381, Locked / Sealed Valve Check List, section l
3.4.3.c stated that manual safety related valves that inter-connect I
redundant trains of safety related systems should be considered for locking devices. These EDG valves were not included in the locked valve program.
The licensee reevaluated the need to lock the identified valves, and
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revised SP-381, Locked / Sealed Valve Checklist, Revision 50, dated 12/2/92, to include provisions to lock the valves. The inspector
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reviewed the procedure and verified the valves were locked in the closed
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position. This item is closed.
7.
Exit Interview The inspection scope and findings were sumarized on February 23, 1993, i
with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in detail the inspection results listed below. Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
Item Number Status Description and Reference VIO 50-302/91-14-01 CLOSED Failure to take prompt corrective i
action to control all control
complex habitability penetrations as required by 10 CFR 50, Appendix B,
Criteria XVI. (paragraph 6.a)
VIO 50-302/91-24-01 CLOSED Failure to perform channel
[
functional tests of anticipatory reactor trips prior to entering Mode l
1. (paragraph 6.b)
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VIO 50-302/92-10-01 CLOSED Failure to perform required review and approval of WR instructions.
(paragraph 6.c)
VIO 50-302/92-12-01 CLOSED Failure to take corrective action to prevent recurrence of a posted high radiation boundsry area door left full open and unmonitored.
(paragraph 6.d)
URI 50-302/92-16-02 CLOSED Development and implementation of corrective action plan for Problem Report 92-0031. (paragraph 6.e)
.
URI 50-302/92-18-02 CLOSED Locking of EDG support systems
cross-connect valves in the closed position. (paragraph 6.f)
LER 91-11 CLOSED Discovery of condition having
,
potential to compromise long term
cooling post-LOCA. (paragraph 5.a)
i LER 92-05 CLOSED Emergency diesel generator could
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exceed maximum rating for certain l
MVP operational lineups die to not
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analyzing the condition of loss of offsite power coincidental with i
engineered safeguards actuation
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during surveillance testing.
(paragraph 5.b)
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LER 92-12 CLOSED Inadequate procedure results in isolation of residual heat removal
system. (paragraph 5.c)
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8.
Acronyms and Abbreviations
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AI
- Administrative Instruction l
- Babcock & Wilcox CFR
- Code of Federal Regulations CP
- Compliante Procedure
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- Decay Heat Removal System t
- Emergency Diesel Generators EN
- Event Notification
- Event Notification System
- Engineered. Safeguards
F
- Fahrenheit FPC
- Florida Power Corporation FSAR - Final Safety Analysis Report gpm
- gallons per minute HPI
- High Pressure Injection i
- Instrumentation and Control j
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,
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=
,
-
.
.
KW
- kilowatt
LCO
- Limiting Conditie:. for Operation
LER. - Licensee Event Report
.
LOCA - Loss of Coolant Accident i
- Modification Approval Record
.
- Motor Operated Valve MVP
- Make-up Pump NRC
- Nuclear Regulatory Commission
OP.
- Operating Procedure
~
p.m.
- post meridiem v
- Plant Review Committee psig - pounds per square inch gauge
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- Radiation Control Area
,
.
- Reactor Protection System RWP
- Raw Water Pump l
SP-
- Surveillance Procedure
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TS
- Technical Specification URI
- Unresolved Item V
- Volt l
- Work Request
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