IR 05000369/1998006
| ML20236P209 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 06/25/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236P201 | List: |
| References | |
| 50-369-98-06, 50-369-98-6, 50-370-98-06, 50-370-98-6, NUDOCS 9807160202 | |
| Download: ML20236P209 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos: 50-369. 50-370 License Nos: NPF-9, NPF-17 Report No: 50-369/98-06. 50-370/98-06
- Licensee: Duke Energy Corporation Facility: McGuire Nuclear Station. Units 1 ar.d 2 Location: 12700 Hagers Ferry Road Huntersville, NC 28078 Dates: April 19. 1998 - May 30, 1998 Inspectors: S. Shaeffer, Senior Resident Inspector M. Sykes Resident Inspector M. Franovich, Resident Inspector R. Gibbs Regional Inspector (M1.2)
W. Stansberry, Regional Inspector (S1.3. S8.1)
Approved by: C. Ogle. Chief. Projects Branch 1 Division of Reactor Projects Enclosure
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9807160202 900625 PDR ADOCK 05000369 G PDR
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EXECUTIVE SUMMARY McGuire Nuclear Station. Units 1 and 2 NRC Inspection Report 50-369/98-06, 50-370/98-06 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covered a six-week period of resident inspection one regional inspection in the area of Maintenance Rule applicability to the reactor protection system, and one inspection in the area of securit Doerations
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Operator cognizance of Unit 1 critical parameters during the Unit 1 refueling outage shutdown were considered goo (Section 01.1)
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Sampled clearances related to Unit 1 shutdown activities were properly pre]ared, authorized, and the components were in the required positions wit 1 the appropriate tags in place. (Section 01.2)
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Licensee actions to address an NRC identified concern regarding procedural guidance for coping with a potential tube break in a reactor coolant pump thermal barrier were appropriate. (Section 03.1)
. The completed upgrade of the control room and simulator control room provided the operators an improved environment for monitoring operation of the units. The implementation was accomplished in a safe manner with limited impact ori operators during transition periods. (Section 05.1)
. The Nuclear Safety Review Board meetings provided good oversight of the facilities operation. Site management gave a realistic presentation of the McGuire site performance to the Board. (Section 07.1)
Maintenanc . Maintenance Rule monitoring of the reactor protection system was in accordance with the guidance of NUMARC 93-0 (Section M1.2)
. Immediate corrective actions to repair an inoperable 1A emergency diesel !
generator following a valid failure on May 19, 1998, were adequate. The {
initial engineering operability evaluation of all four station diesels was acceptabl An inspector followup item (IFI) 98-06-01, Root Cause and Corrective Actions for Failure of the 1A emergency diesel generator number 6R Cylinder Exhaust Valve Seat, was identified pending review of the metallurgical examination results, root cause report, subsequent !'
corrective actions, and review of previous Duke quality assurance audits of approved vendor (Section M2.1)
. System engineering response to a potential common surveillance problem identified at the Catawba site was proactive, immediate. and provided assurance of continued operability of the auxiliary building and other safety-related ventilation systems at McGuire. (Section M7.1)
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Enaineerina
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A comprehensive and very detailed safety evaluation was performed by the corporate nuclear engineering department to evaluate suitability of new fuel that contained cladding contaminated with sodium chloride. The licensee's and fuel vendor's approach for addressing potential Nacl effects on fuel performance included a rigorous examination of the issue with good consultation by independent industry fuel experts. (Section E2.1)
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The licensee's efforts to minimize the likelihood of a dual unit trip similar to that experienced on September 6. 1997. by performing electrical loading modifications was prudent and should have a positive effect on plant reliabilit (Section E3.1)
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An Unresolved Item was identified to further evaluate inspection requirements for the refueling water storage tanks and review documentation of the interior tank coatings used to support continued
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operabilit (Section E4.1)
. The licensee's initial documentation of the problem within PIP 1-M98-0249 was weak in not providing a thorough operability evaluation for the refueling water storage tank. Additional problems were noted in the i initial mis-identification of the material used in the construction of !
the refueling water storage tank and misunderstanding of actual past '
inspections performed en the tank. (Section E4.1)
Plant Supoort i
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. The licensee adequately complied with the access authorization and )
control of personnel criteria of the Duke Power Company Nuclear Security I and Contingency Plan Procedures. The licensee acted conservatively in reporting an access authorization issue to the NRC. (Section S1.3)
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Report Details Summary of Plant Status Unit 1 Unit 1 began the inspection period at approximately 100 percent power. On May 29 the unit began a controlled shutdown to begin the planned end of cycle 12 refueling outage. At the end of the inspection period, the unit was in cold i shutdown, preparing-for defueling operation ]
Unit 2 Unit 2 operated during the inspection period at approximately 100 percent powe Review of Uodated Final Safety Analysis Report (UFSAR) Commitments q
While performing inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that were related to the areas inspecte i The inspectors verified that the UFSAR wording was consistent with the f
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observed plant practices. procedures, and parameter I. Ooerations l 01 Conduct of Operations 01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent i reviews of ongoing plant operations. In general, the conduct of operations was professional and safety-conscious. Operator cognizance of Unit 1 critical parameters during the Unit 1 refueling outage shutdown was considered goo Specific events and noteworthy observations are detailed in the sections which follo .2 Ooerations Clearances (71707)
The inspectors reviewed a number of clearances related to Unit 1 shutdown activities during the inspection period. The inspectors observed that the clearances were properly prepared and authorized. and that the tagged components were in the required positions with the appropriate tags in plac Operations Procedures and Documentation
03.1 Emergency Procedure Adeauacy for Cooina with a Tube Ruoture in the ;
Reactor Coolant Pumo (RCP) Thermal Barrier Heat Exchanaer l Insoection Scooe (71707)
Inspectors reviewed the licensee's response to an NRC identified concern regarding adequacy of emergency operating procedures for coping with a l
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potential rupture of the reactor coolant pump thermal barrier heat exchanger, Observations and Findinas )
i The inspectors' concerns regarding a potential 285 gallons per minute l (gpm) interfacing system loss of coolant accident (LOCA) through the RCP i thermal barrier heat exchanger were documented in IR 97-20. Problem I Investigation Process (PIP) report 0-M-98-1529, identified licensee corrective actions to revise EP/1.2/A/5000/ECA-1.2. LOCA Dutside i Containment. to include appropriate steps to identify and isolate a !
ruptured thermal barrier. In addition, the corresponding annunciator
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response procedure OP/1.2/A/6100/010K will be improved to provide o)erators with additional guidance to co]e with the subject scenari Tlese corrective actions are planned to ]e incorporated in the next few month c. Conclusions
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The inspectors concluded that the proposed corrective actions to enhance procedure adequacy for coping with a potential tube break in an RCP ,
thermal barrier were appropriate. IFI 50-369.370/97-20-03 is close Operator Training and Qualification 05.1 Comoletion of Control Room and Simulator Work Station Unarade ,
a. Insoection Scooe (71707)
The inspectors observed activities associated with the upgrade of the main control room and simulator to evaluate the impact on plant operations and the effectiveness of the implementation proces Observations and Findinas The inspectors monitored the licensee's upgrade of the McGuire main control room and the simulator control room which began in Februar l 1997. The upgrade was accomplished in conjunction with the steam generator replacement project (SGRP) outages on both units which included the replacement of the units' operator aid computers and digital rod position indication systems. The licensee chose the SGRP outages to implement this upgrade to limit impact on an operational ,
unit. The upgrade included replacement of the reactor operator (RO) and l senior reactor operator (SRO) work stations, incorporation of new video screens for operator information and logging, the addition of increased use of control board mimics, and more distinctive work area locations to allow for improved control of control room traffic. During the implementation of activities associated with the control room modifications, the inspectors did not observe any adverse impacts to safe operation of the units. The simulator control room was u) graded in advance of the main control room to allow for operator feedbacc and acclimatio , _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - . . -
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The inspectors considered that the u) grade allowed for improved monitoring of the units by placing t1e operators in positions more prone to facing the control boards and elevating the control room shift manager to provide better overview of control room activities. Other attributes were im) roved desk layouts providing better immediate access to emergency and a] normal procedures, as well as better control of routine task Conclusion The inspectors concluded that the comaleted upgrade of the control room and simulator control room provided t1e operators an improved environment for monitoring operation of the units. The implementation was accomplished in a safe manner with limited impact on operators during transition period Quality Assurance in Ooerations 07.1 Nuclear Safety Review Board (NSRB)
I Insoection Scone (4500)
The inspector attended the NSRB meeting to assess how the licensee evaluated overall plant performance and responded to plant issue Observations and Findinas On May 14. 1998, the inspector attended the McGuire portion of the NSRB meeting held at the Catawba site. Site presentations to the board included plant performance, reportable events, violations, trends, areas for improvement, and other relative issues. The inspector considered that t1e information presented to the NSRB gave a realistic view of overall plant performance. Numerous proposals for improved performance were suggested and documented for resolution, Conclusions The inspector concluded that the NSRB members provided good insights to plant management for potential improvements. McGuire site management gave a realistic presentation of the McGuire site performance to the Boar Miscellaneous Operations Issues 0 (Closed) Licensee Event Report (LER) 50-369/97-03: Inoperability of Both Trains of the Control Room Ventilation System Due to Isolated Air Intake Valves This event involved inoperability of the common main control room ventilation system (VC) during a planned maintenance activity on train B of VC. The B train of VC was out of service due to the associated outside air intake valves being closed to support maintenance on
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radiation monitor 1 EMF-43 During the planned activities, the operators received a high radiation alarm on radiation monitor 1 EMF-43A and took actions to isolate the remaining train A outside air intakes as directed by the applicable alarm response procedure. This action resulted in both trains of VC becoming inoperable for a period of approximately two minutes until the 1 EMF-43A alarm cleared, allowing the intakes to be reopened. The cause of the Train A radiation alarm was later attributed to spurious operation, which was confirmed by increased radiation sam] ling. The inspectors reviewed the operators immediate response to t1e event and determined the actions appropriate. In addition, the inspectors reviewed additional corrective actions identified in the subject LER and concluded that the licensee responded appropriately to the event. Safety im3act to the shutdown and operational unit was minimal. This LER is close .2 (Closed) LER 50-369.370/97-04. Revision 0 and 1: Main Steam Safety Valve Technical Specifications (TS) Inaccuracies The inspectors evaluated the licensee's actions following the-
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identification of non-conservative TS requirements of TS 3.7.1.1.a and Table 3.7-1 outlining the maximum allowable power range neutron flux high setpoint with inoperable steam line safety valves during four loop operation. The TS allows continued power operation at less than 43 percent rated thermal power with 3. main steam safety valves inoperabl This TS was determined to be non-conservatism by the licensee following a detailed analysis. The licensee established administrative controls to prohibit operation with three or more steam line safety valves inoperabl The licensee has also submitted a formal TS amendment request to correct this non-conservatism. The inspectors reviewed the instructions provided to the operating crews and confirmed that the instructions were ap3ropriate to prevent operation outside plant design basis. Therefore, t1is item is close .3 (Closed) LER 50-369/97-08. Revision 0 and 1: Inoperability of the Auxiliary Feed Water (AFW) System Due to Potential Air Entrainment l
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On June 11, 1997, the licensee submitted an LER abstract describing potential air entrainment from the non-safety related AFW suction sources based on preliminary calculations. After further analysis, the licensee determined that no operability concerns existed. Therefore, this LER is closed; however, inspector followup item (IFI) 97-08-0 Potential Airbinding of AFW Pumps, will remain open pending further NRC review of the licensee's supporting calculations.
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08.4 (Closed) IFI 50-369.370/97-20-03: Emergency Procedure Adequacy for j Coping with a Tube Rupture in the RCP Thermal Barrier Heat Exchange !
. This IFI is closed (reference section 03.1 of this report)
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l 5 II. Maintenance M1 Conduct of Maintenance M1.1 GSneral Comments
_ Insoection Scone (61726 and 62707)
The inspectors observed portions of the following work activities:
Procedure / Work Order Title PT/2/A/4206/001B 2A Safety Injection Pump Testing
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98028823 2A Hydrogen Analyzer Quarterly Preventive Maintenance (PM) and Calibration b. Observations and Findinas The inspectors witnessed selected surveillance tests to verify that approved procedures were available and in use: test equipment was calibrated: test prerequisites were met: system restoration was completed: and acceptance criteria were met. In addition the inspectors reviewed or witnessed routine maintenance activities to verify, where applicable, that approved procedures were available and in use, prerequisites were met, equipment restoration was completed, and maintenance results were adequat c. Conclusion The inspectors concluded that these and other routine surveillance and ,
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maintenance activities were completed satisfactoril {
M1,2 Maintenance Rule Monitorina of the Reactor Protection System )
, Insoection Scope (62706)
This inspection was conducted to review monitoring of the reactor protection system as a result of problems identified at other utilities l
during Maintenance Rule baseline inspections. Two areas of concern evolved as a result of those inspections: First, there was concern that 1 the system was not being monitored at a low enough level to prevent masking of problems, which would eventually result in a plant trip or an inadvertent safety system actuation. Second, it was found that some
, utilities had classified the system as risk-significant, and were not i monitoring unavailability in accordance with NUMARC 93-01 guidance. The inspector reviewed the system performance criteria, Maintenance Rule and non-Maintenance Rule data, system surveillance testing, block and logic l diagrams, and interviewed the system engineer and the Maintenance Rule l coordinator to determine if the monitoring of the reactor protection l system under the Maintenance Rule was in accordance with NUMARC guidance.
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b. Observations end Findings The licensee's performance criteria and Maintenance Rule data collection was established to monitor reliability at the train leve However, additional discussion with the system engineer determined that system monitoring, in many cases, went to a far lower component level. This was evidenced by the fact that the system engineer reviews all work orders on the system and Jersonally resolves most of the deficiency reports on the system. T1e licensee also trends card fai'.ures; power supply failures and drift; detector current. insulation resistance, and capacitance; solid state system failures and logic board failures; and
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reactor trip breaker timings, trip bar force, under voltage trip attachment forces and under voltage coil dropout voltages. Data supporting this m'nitoring reviewed by the inspector provided clear evidence of detailed monitoring at the component leve {
The second concern at other utilities related to unavailability monitoring for risk-significant portions of the reactor protection ,
system. All portions of the McGuire system were classified as non-risk j significant with the exception of the reactor trip breakers which were J classified as risk-significant. The inspector verified that this j classification was in accordance with the licensee's robabilistic risk assessment. The inspector determined that unavailabi ity was not being !
monitored on the breakers, and questioned the licensee concerning this l a] parent deviation from the NUMARC guidance. The licensee stated that i t1e only possible unavailability would be during testing of the trip breakers. This testing was required by technical specifications to be accomplished once every 62 days, one train at a time, and limited to a L two-hour duration. In a typical year, this would amount to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of )
unavailability, which was bounded by the licensee's probabilistic risk assessment. In addition, the testing was accomplished with both the l reactor trip breaker, and the bypass breaker in the " racked in" position. With the breakers in this position, the manual trip actuation was not defeated, and one immediate operator action (initiation of a reactor manual trip) would restore the breaker safety function, and, in fact, would initiate the function. As a result, the licensee did not consider the breaker to be unavailable during testing, and had established the breaker performance criteria as 100% availability and zero-functional failures. The inspector verified that condition monitoring was being conducted on the breakers as a result of these performance criteria, and the NRC agreed with the licensee's position L concerning unavailability.
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L M2' Maintenance and Material Condition of Facilities and Equipment ;
' M2.1- Unit 1 Emeraency Diesel Generator Valid Failure Durina Monthly Surveillance j, J_nsoection Scoce (62707. 40500)
L The inspectors reviewed the facts and circumstances of a failure of the l 1A emergency diesel generator (EDG) during a monthly operability test.
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EDG diagnostic data, the Nordberg EDG vendor manual, vendor 3rocedures for' cylinder head rebuild, work order number 98049165. and tle post- l maintenance operability test were.. reviewed. The inspectors focused on i the potential for a common-mode: failure since all four McGuire station EDGs had cylinder heads recently rebuilt offsite by a vendor. The ~l l inspectors also focused on current operability, industry operating j l
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experience; valve and seat installation techniques and the licensee's quality assurance process for vendor process oversigh ,
1 Observations and'Findinas '
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l Two Nordberg FS-1316HSC EDGs are provided for each unit for onsite l P emergency alternating current (AC) power. Each EDG is rated for 4,000 i kilowatts and is-designed with sufficient capacity-to supply ac power j l for the recuired loads to safely shutdown the unit'following a design l L basis accicent. Each EDG is a 4 stroke. 16 cylinder engine with an j l intake valve and exhaust valve for each cylinder head. All four EDGs l were refurbished last year with the cylinder heads rebuilt, offsite, by l NAK Engineerin j On May 19, 1998, routine. monthly the Unit 1 EDG operability test. 1ATh ex!erienced 6R cylinderaexhaust failure during tem)erature was below normal and not firing properly. A failure of the ex1aust ;
valve seat.had occurred with a 120 degree arc section of the valve seat (
broken away. Maintenance personnel retrieved the broken piece from the turbo-charger ring catcher. Metallurgical examination was in-progress at the end of the inspection period to determine the failure mode. The valve seat, which was com)osed of a tungsten molybdenum steel alloy, was imbedded in the cylinder lead. Although 15 of the 16 cylinders carried the full load, the licensee. considered this a valid failure due to increased stress on the crankshaft, which could lead to catastrophic l failure. This was the first failure of the 1A EDG in over 100 run although less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> had been logged on the engine since the
. major overhaul. The ins)ectors concluded that the 1A EDG performance
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remained within established target goals for EDG reliability. Howeve IFI 98-06-01, Root Cause and Corrective Actions for Failure of the 1A EDG number 6R Cylinder Exhaust Valve Seat, was identified pending NRC review of the metallurgical examination results, root cause report,
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subsequent. corrective actions, and review of previous. Duke quality assurance audits of NAK Engineering i
The ins)ectors verified that the licensee appropriately declared the EDG !
inopera)1e at the time'of discovery on May 19, 199 On May 20.-1998, j
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the 6R cylinder head was replaced and an operability test was satisfactorily performe The inspectors were concerned that potential damage during removal from the cylinder head may have occurred if valve seats were reused during the rebuild. Inspectors' review of the Nordberg vendor manual indicated that removal of inserts. may be performed using 0.75 inch spot welds at 120 degrees apart. This appeared to coincide with the failure orientation. However, the system engineer indicated that McGuire specified that new valve . seats were to be used in the rebuil The licensee's preliminary investigation revealed that the seat was cocked and the acre in the head may not have been completely roun Inspectors questioned if the Nordberg vendor manual recommended installation practices were used by NAK. by letter dated May 21, 199 NAK Engineering notified McGuire that the Nordberg vendor manual, procedures, and McGuire station procedure for cylinder corrective maintenance were being utilized. However, the inspectors did observe that NAK did not document actual measurements (shrink fit dimensions) of
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each bore hole and valve seat insert for each cylinder hea Additionally, during the review of this event the inspectors also noted that the local alarm panel annunciator response procedure for hi/lo exhaust temperature did not identify a failed seat as a potential caus Performance data gathered during a post-rebuild break-in run in 1997 revealed lower than normal 6R cylinder firing pressure during low load conditions. However, the 6R had normal firing pressure at full load conditions. The licensee had concluded that the full load firing pressure was usually an important factor where engine balance is of concern, and therefore this low load anomaly was dismissed during the review of break-in data. Following the seat failure, the licensee postulated that the lower firing pressure may have been attributed to gas leakage between the seat insert and the bore in the head. At increased load, the licensee believed that high temperatures could have reduced the gap and help explain why firing pressure was consistent with the other 15 cylinders. Prior to the failure, the subject cylinder continued to meet acceptance criteria during tests conducted between the 1997 rebuild break-in run and the May 19, 1998 test. Inspectors review of engine diagnostics from the break-in runs confirmed no similar low-temperature anomalies on the other three diesels and the other 15 1A EDG cylinders for the full range of load condition c. Conclusions The immediate corrective actions to repair an inoperable 1A emergency diesel generator that experienced a valid failure on May 19, 1998, were adequat The initial engineering operability evaluation of all four station EDG appeared reasonabl However IFI 98-06-01. Root Cause and Corrective Actions for Failure of the 1A EDG number 6R Cylinder Exhaust Valve Seat, was identified pending NRC review of the metallurgical examination results, root cause report, subsequent corrective actions,
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and review of previous Duke quality assurance audits of NAK Engineerin M7 Quality Assurance in Maintenance l
M7.1 Resoonse to Catawba Potential Technical Specification Surveillance Inadeauacy In:cettiori Scooe (71707)
The inspectors monitored the licensee's immediate follow up to PIP report 0-C98-1817 identified at the Catawba Nuclear Station for pott ial impact on the McGuire unit Observations and Findinas The subject Catawba PIP involved a potential missed auxiliary building TS surveillance due to the system's moisture separators not being included in the delta-P testing for the components. The inspectors reviewed the McGuire ap)licable TS and discussed the Catawba PIP with
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the system engineer witlin one day of the PIP being received at McGuir The inspectors noted that the system engineer was aware of the Catawba PIP and had already determined that the specific problem did not exist at McGuire due to specific equipment differences. The involved system engineer also reviewed other plant ventilation systems for similar concerns. No problems were identified. However, in the annulus venti 4 tion system. several bar type heaters were evaluated for potential flow blockage impact and found to be acceptabl Conclusion The inspectors concluded that system engineering response to a potential common surveillance problem at the Catawba site was aroactive, immediate, and provided assurance of continued opera)ility of the auxiliary building ventilation system at McGuir M8 Miscellaneous Maintenance Issues M (Closed) VIO 50-369/97-08-02: Inadequate Procedure for Performing Analog Channel O r ational Test (ACOT)
(Closed) LER 50-369/97-07: Unit 1 Reactor Trip During MODE 3 Power Range Instrumentation Calibration On ?iay 14. a Unit 1 reactor trip and feed water isolation occurre The ,
reactor trip occurred c , 2 out of 4 logic for turbine trip / reactor trip l during power range instrumentation calibration. Prior to the event the l reactor was in Mode 3 with shutdown banks A and B withdrre. RCS boron diluted, and a stable reactor temperature of approximately 557 Station personnel were 3reparing for an approach to criticality and subsequent zero power p1ysics testing. As part of the are)arations, the licensee entered TS 3/4.10.3 "Special Test Exce) tion" w1ici requires, in part, that each power r6 age channel shall be su) ject to an Analog
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Channel Operational Test (ACOT) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating startup and physics tests. During the ACOT of N-42. PR channel N-41 (connectea to the reactivity computer) was improperly bypassed due to an inadequate ACOT procedure that resulted in a sub-critical reactor tri The inspectors reviewed the LER, the licensee response to the violation, the root cause analysis, and the corrective actions which included a revision to station procedures for ACOT of power range channels with appropriate guidance to bypass an inoperable PR channel for the startup configuration. The inspectors concluded that these changes were acceptable: therefore, this violation and related LER 50-369/97-07, Unit l 1 Reactor Trip During MODE 3 Power Range Instrumentation Calibratio are close III. Enaineerina E2 Status of Engineering Facilities and Equipment
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E2.1 Unit 1 New Fuel - Sodium Chloride (Nacl) Contamination of Zircalov Insoection Scope (37551. 40500)
The inspectors reviewed the licensee's safety evaluation for use of new fuel that was contaminated with low levels of sodium chloride. The inspectors discussed the issues with station personnel, attended a Plant Operating Review Committee (PORC) presentation. reviewed the supporting 10 CFR 50.59 evaluation. and consulted the UFSA !
1 Observations and Findinas ;
Framatome Cogema Fuels (FCF), fuel suppliec for McGuire, and the ,
licensee have investigated the cause and effects of Nacl that was l inadvertently introduced on Zircaloy cladding during manufacturin A )
change in the final tube cleaning process at the cladding supplier (Zircotube of Paimbouef, France) resulted in the contamination of several lots of Zircaloy cladding with Nac The contamination was discovered during a FCF investigation of a high number of fuel rod end plug weld rejects. Twenty five of the new fuel assemblies for Unit 1 operating cycle 13 contained rods contaminated with NaC1, which appeared as water stains ori the interior of the claddin FCF and Duke have performed a root cause analysis. Elevated levels of Nacl were in potable water during the hot water tube rinse. Zircotube now uses deionized water onl Potential fuel failure mechanisms were evaluated by FCF and Duke with consultation by independent industry fuels experts. A comprehensive failure mode and effects analysis (FMEA) was performed to assess impact of Nacl examining stress corrosion cracking, internal hydriding, end plug weld integrity, and impact on heat transfer. Effects of Chlorine were considered insignificant when compared to fission gas attack. No residual water should have remained in the MaCl crystals based on
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physical properties of Nacl and confirmed low humidity conditions during fuel fabrication. Also, concentrations <f Nacl were significantly less than what is permitted by FCF specifications for maximum chloride residue on fuel pellets (25 parts per million). FCF also completed an autoclave corrosion test which confirmed no adverse impact on cladding performance. The inspectors concluded that the 10 CFR S0.59 safety evaluation was satisfactorily com)leted for use of the contaminated fuel. The inspectors discussed t1e issues with NRR reactor systems branch and the NRR Project Manager during a conference call on May 12, 1998. No specific concerns were expressed.
' Conclusions l A comprehensive and very detailed safety evaluation was performed by the corpcrate nuclear engineering department to evaluate sd tability of new fuel that contained cladding contaminated with sodium chloride. The inspectors considered the licensee's and fuel vendor's ap3 roach for addressing potential Nacl effects on fuel performance to )e a rigorous examination of the issue with good consultation by independent industry fuel expert E3 Engineering Procedures and Documentation E3.1 Modification of 120 Volt (V) AC Auxiliary Control Power System Loads Insoection Scooe (37551)
The inspectors reviewed and evaluated pre-implementation plans for modification MGMM-9994 which would remove critical Unit I loads from 120 VAC auxiliary control power system panel board KXB and realign them to panel board KXA. A loss of power to panel board KXA resulted in a dual reactor trip on September 6,1997, since essential loads for both units were being powered from KXA at the tim Observations and Findinos l
The 120VAC Auxiliary Control Power System provides a normal source of power to loads requiring regulated power and provides a normal and l alternate ]ower supply to loads requiring an uninterruptable power supply. T1e auxiliary control power system consists of two 120VAC power l panel boards, KXA and KXB. Panel board KXA carries primarily Unit 1 J
loads and panel board KXB carries primarily Unit 2 loads: however, each l of the panel boards carry some loads from the opposite unit. The I modification was developed to segregate the critical Unit I load; on panel board KXA. The affected loads include Unit 1 Electro Hydraulic Control System and Main Steam Isolation Valve Control. A similar modification was proposed for the upcoming Unit 2 outage to segregate the critical Unit 2 loads on KXB to minimize the likelihood of a similar dual unit trip following a loss of power to either panel board KXA or KX _ - _ _ _ _
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The inspectors evaluated the licensee plan for im] lamentation and noted l that realignment of loads will be performed with Jnit 1 defueled and Unit 2 in MODE 1 to eliminate the potential for a dual reactor trip and consequently reduce the impact on control room operators. The !
inspectors independently reviewed the Unit 2 wort schedule to identify I any plant activities that could increase the potential for a 31 ant I transient during modification implementatio No negative scledule l interactions were identified. As an additional precaution. the licensee I also identified this realignment of loads as a critical maintenance f activity and had identified the necessary management controls and i contingencie ; Conclusions The inspectors concluded that the licensee's efforts to prevent l recurrence of a dual unit trip by performing this modification was j prudent and should have a positive effect on plant reliabilit E4 Engineering Staff Knowledge and Performance
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E4.1 Lack of Insoection and Docursentation for Refuelina Water Storaae Tank l (RWST) Interior Coatir.as ' Inspection Scooe (71707)
The inspectors discussed PIP reports 0-M97-3386 and 1-M98-0249 with licensee engineering personnel concerning past inspections performed on each unit's RWST and qualification / potential deteriorization of the l tanks' interior coating l Observations and Findinas In mid-1997, the inspectors questioned the licensee regarding what l inspections had been performed on the RWST The licensee indicated i that the tanks had been inspected internally; however records of the inspection were not readily available. During a subsequent licensee initiated review of the refueling water system, on September 17. 1997 l the licensee questioned the need for periodic vessel / code related i inspections for the RWSTs. PIP 0-M97-3386 was initiated to document the l basis for periodic inspections for the RWSTs and what inspections had been performed in the pas On January 26, 1998. engineering personnel initiated PIP 1-M98-0249 which questioned the integrity of the Plasite (epoxy type) coating inside the Unit 1 RWST. The PIP description indicated that the material of the Unit 2 RWST appeared to be stainless steel with no Plasite coating. The PIP was initiated to determine the condition of the Unit 1 RWST interior coatings, based on a lack of information available regarding its long term qualification. The PIP further indicated that design engineering correspondence indicated that the service life of the coating (Plasite 7155) was identified as 10 years. The McGuire RWSTs have been in service for over 17 year _ _ - _ _ _ _ _
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During the current inspectica period, the inspectors met with cognizant engineering personnel to discuss the above PIP During this meeting, the inspectors were informed that the licensee had recently identified that both of the RWSTs were fabricated from carbon steel with Plasite interior coatings and that neither tank had any record of internal inspection. Based on the above, the inspectors questioned the operability of the tanks considering the lack of service life I qualification information regarding the RWST interior coatings and the lack of visual inspections to verify the coating integrit The inspectors further questioned if the RWST interior coatings deteriorated, could they pose a threat to the containment sump during a LOC PIP 1-M98-0249 documented an operability review that analyzed potential corrosion rates of the RWSTs' carbon steel in a boric acid environmen This review concluded that the RWSTs were operable based on negligible impact of the anticipated corrosion on the seismic capability of the tanks to perform their intended safety function. In addition, the PIP referenced past degradation of the Oconee facility boric dcid tank This information was used to support continued McGuire operability. The inspectors noted that the PIP operability evaluation did not address the potential degradation of the coatings and their potential effect on the containment sump or post-accident fuel performanc The licensee concluded that the RWSTs interior coating were not degradad based on their avaluation and that routine sampling of the tanks did riot reveal any abnorrr.alities. At the end of the ins 3ection period, the licensee had determined that no ins)ections of t1e Unit 1 RWST interior coating would be performed during t1e current rO nng outage: although future inspection were being considered. Unres med Item 50-36 /98-02, Refueling Water Storage Tank Interior Coating Inspection, was identified to further evaluate inspection requirements for both refueling water storage tanks and documentation for the interior tank coatings to support continued operabilit Conclusion An Unresolved Item was identified to further evaluate inspection requirements for the refueling water storage tanks and review documentation for the interior tank coatings used to support continued operabilit The licensee's initial documentation of the problem within PIP 1-M98- l 0249 was weak in not providing a thorough operability evaluation for the l refueling water storage tanks. Additional problems were noted in the l initial mis-identification of the matericl used in the construction of the refueling water storage tank and misunderstanding of actual past inspections performed on the tan _ _ _ _ - _ _ _ _ _
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E8 Miscellaneous Engineering Issues E8.1 (Closed) URI 50-369.370/96-06-02: Refueling Practices Involving Full Core Off Load Into SFP. Unresolved Item 96-06-02 was opened to evaluate licensee's past 3ractice of offloiding the full core on a routine basis with regard to t1e licensing basis for normal refueling practices. As discussed in IR 50-369/96-06, the staff reviewed licensing basis regarding spent fuel pool decay heat removal and refueling outage core offload practice In a revision of the UFSAR (Section 9.1.3) dated January 12, 1995, the licensee describes the heat loads assumed for analyzing the spent fuel pool cooling section. The UFSAR states:
Normal Heat Load: Assumes one-third core has been placed in the pool seven (/) days after shutdown. The remainder of the pool, less 193 spaces, is filled with previous McGuire discharges from normal refueling operations and Oconee spent fuel which has decayed at least five (5) years. The 193 empty spaces are reserved for a full core discharg '
Abnormal Heat Load: Assumes one full core discharge consisting of ;
three batches. The batches are irradiated 23.5 days, one-yea and two years respectively. In addition, one refueling batch has decayed 36 day The remainder of the pool is filled with previous McGuire discharges from normal refueling operations and Oconee spent fuel which has decayed at least five (5) year Further. UFSAR Table 9-5 describes the calculated bulk spent fuel temperature for these two cases (identified in the Table as the normal maximum and abnormal maximum heat loads) with two spent fuel pool cooling system configurations (one and two trains operating) analyzed for each case. An assumed heat load. [ spent fuel pool] design basis tem)erature and calculated spent fuel pool temperature were provided for eac1 cas The normal maximum heat load case, analyzed with one cooling train operating, results in a calculated spent fuel pool temperature of 133 The abnormal maximum heat load case, analyzed with one cooling train operating, results in a calculated spent fuel pool temperature of 178 F. In a footnote to Table 9-5. the licensee stated, regarding design basis for spent fuel pool temperature, that 140 F was used as a maximum for [ spent fuel pool] structural calculation Tne inspectors reviewed records from past refueling outages and confirmed that:
(1) The licensee has offloaded the full core during each refueling since plant startu (2) During all past full core offloads, the licensee did not offload the full core prior to the 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown assumed in the UFSA _
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(3) The inspectors did not review records of spent fuel 1001 cooling system operability records, but did determine that t1e maximum temperature reached following a full core offload was 120 F. This temperature was well below the design criteria of 140 F described in the UFSA The inspectors observed that UFSAR Section 9.1.3.1 by introducing the terms normal and abnormal, characterizes each of the design basis cases in a way that could imply that full core offloads are to occur on a less frequent basis than partial core offloads. The inspectors were unable to conclude: however, that Section 9.1.3.3. represents a specific commitment to limit the frequency with which full core offloads are conducted. As noted above, nothing in the UFSAR description of the s)ent fuel pool cooling design basis is sensitive to the frequency with w1ich full core offloads are conducted. Therefore, the inspectors concluded that the practice of offloading the full core during each refueling outage did not represent a change to the facility or a change to the procedure described in the UFSAR and thus did not require a review pursuant to 10 CFR 50.59. On this basis URI 50-369/96-06-02. is
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close IV. Plant Sucoort R1 Conduct of Radiation Protection and Chemistry R1.1 General Comments (71750)
The inspectors made frequent tours of the controlled access area and reviewed radiological postings and workar adherence to protective clothing requirements. Locked high radiation doors were properly controlled, high radiation and contamination areas were properly posted, and radiological survey maps were updated to accurately reflect radiological conditions in the respective area S1 Conduct of Security and Safeguards Activities S1.3 Access Control-Personnel Insoection Scooe (81070)
The inspectors evaluated the licensee's access control of personnel concerning an April 13. 1998, security badge event. This evaluation was to ensure that.the licensee implemented criteria in the Duke Nuclear Security and Contingency Plan (PSP). and appropriate corporate and Security Plan Procedures (SPPs). Observations and Findinas On February 25, 1998, a Qualified Maintenance Support Specialist. In (OMSS) employee notified his management that he would be leaving their employment to become a contract employee for another Duke Power
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organization. Duke Engineering and Services (DE&S), on April 1,199 _ _ _ - _ . - _ _
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On March 31, 1998, QMSS favorably terminated the employee without-notifying McGuire Security; consequently, his site access to Duke nuclear facilities remained vali On April 13. 1998. the Access Control Group at the General Office received a letter from QMSS saying that they had terminated an em)loyee on March 31, 1998, and no longer required unescorted access to Duce Power Company nuclear facilities. Review of the access control database showed that employee still had an active security access badge at McGuire Nuclear Station. The General Office notified McGuire site security at approximately 11:45 a.m. , of the situation and immediately terminated the employee's site access. At 5:45 p.m., on the same day, the site Access Control Supervisor learned from the terminated employee that he had entered the prutected area on April 7 to keep his security badge from being placed on Administrative Hold because of ten-day non-use. With no malevolent intent, the employee entered the protected area, walked around the area for eight minutes, and then existed, without entering any vital areas that his badge could have allowed him to enter. Because entry into the protected area by a terminated
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employee is a one-hour reportable event under 10 CFR 73.71, the licensee notified the NRC Operations Center and Senior Resident Inspector of the event at 6:20 The root cause of this event was the lack of coordination and communication between involved badge and job s]onsors. The terminated employee had the same McGuire Nuclear Station .3adge s3onsor and job sponsor for both of the OMMS and DE&S projects. The Jadge sponsor was unaware that the employee had changed employment. The employee needed unescorted access to the site as determined by the March 31 day badge review, which was performed by the badge sponsor. The badge sponsor had even extended the employee's badge expiration date from May 31, 1998, to April 1, 1999 on March 30, 1998, due to being assigned to the new project. The job sponsor was aware of the employer change, but was unaware that changes needed to be made to the employee's employment status. The employee continued to work on the OMSS project after changing employment, while training for the new project, and was being paid after March 31, 1998 by DE& During the Problem Investigation Process (PIP) (0-M98-1118), the licensee determined that the employee was transferred from one Duke Power employer to another rather than terminated and rehired. The licensee retracted the NRC Notification (34067) on April 16, 199 Duke Power Nuclear Policy Manual-2. Nuclear System Directive B.1 Access Restrictions states, "If the termination of employment is VOLUNTARY / FAVORABLE the individual's Security Badge must ce restricted when there is no longer a need for work related unescorted access." In i this situation, the employee met the following requirements:
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Approved unescorted access authorization per Duke Power Nuclear Access Authorization Progra _ _ _ _ _ - - _ _
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Fingerprinte Fitness for Duty Progra Current General Employee Trainin The Pispectors' evaluation revealed the following:
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that the licensee maintained the requirements for authorized unescorted access to the McGuire Nuclear Station before, during and after the employee transferred from QMSS to DE& that the initial one-hour call to NRC Operations was timely and justified based upon the facts known at the time of the cal that the subsequent call to cancel the one-hour notification was appropriat Conclusion f
The licensee adequately complied with the access authorization and control of personnel criteria of the Duke Power Company Nuclear Security and Contingency Plan, and appropriate corporate and Security Plan Procedure The licensee acted conservatively in reporting this matter to the NR S8 Miscellaneous Security. and Safeguards Issues 5 (Closed) Unresolved Item 50-369.370/97-13-02: The Licensee Failed to Deactivate and/or Deny Protected Area Access to Terminated Employee (Closed) Unresolved Item 50-369.370/97-13-03: Licensee Failed to Control Protected Area Access Badges. In That The Badges Were Taken Outside The Protected Area Unescorted By Security Personne On September 11, 1997, a description of a3 parent violations was provided to the licensee for the issues described )y URO 50-369.370/97-13-02 and 97-13-0 No EEIs were issued and EA Case Number 97-411 was assigne On September 19, 1997. a predecisio'lal enforcement conference for EA Case Number 97-411 was held with the licensee in attendanc Following the conference, a Notice of Violation was issued on September 26. 199 Based on the NOV issued, the URIs above are closed and the violations identified in the above NOV will be tracked as:
VIO EA97-411/01013 The Licensee Failed to Follow Procedures for Security Badges and Access Control.
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l VIO EA97-411/01023 The Licensee Failed to Follow j
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Procedures for Security Badges and {
Access Contro !
, VIO EA97-411/02014 The Licensee Failed to Follow Security Procedure EXA0-2 on Ten i Occasions Between The Dates of .
February 5. 1997 and July 10. 1997, j When The Licensee Failed to Control Protected Area Access Badges. and The Badges Were Taken Outside The Protected Are .
V. Management Meetinos j X1 Exit Meeting Summary I The resident inspectors ) resented the inspec' ion results to members of
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licensee management at t1e conclusion of the inspection on June 2. 1998. The licensee acknowledged the findings presented. No proprietary information was identifie PARTIAL LIST OF PERSONS CONTACTED Licensee Barron. B., Vice President. McGuire Nuclear Station Bhatnagar, A., Superintendent. Plant Operations Boyle, J., Civil / Electrical / Nuclear Systems Engineering Byrum. W., Manager Radiation Protection Cash M., Manager, Regulatory Compliance Dolan B., Manager. Safety Assurance Evans W., Security Manager Geddie. E., Manager, McGuire Nuclear Station Peele. J. , Manager. Engineering Louck L., Chemistry Manager Pederson. T. . Maintenac :e Rule Coordinator Peele J. , Manager. Engineering Thomas, K.. Superintendent. Work Control Travis, B., Manajer. Mechanical Systems Engineering INSPECTION PROCEDURES USED IP 71707: Conduct of Operations IP 62706: Maintenance Rule IP 62707: Maintenance Observations IP 61726: Surveillance Observations i
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IP 40500: Sel f-Assessment IP 37551: Onsite Engineering IP 71750: Plant Support IP 81070: Conduct of Security and Safeguards Activities s
ITEMS OPENED. CLOSED AND DISCUSSED OPENED EA97-411/01013 VIO The Licensee Failed to Follow Procedures for Security Badges and Access Control (Section j S8.1)
EA97-411/01023 VIO The Licensee Failed to Follow Procedures for Security Badges and Access Control (Section 58.1)
EA97-411/02014 VIO The Licensee Failed to Follow Security Procedure
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EXAO-2 on Ten Occasions Between The Dates of February 5. 1997 and July 10. 1997 When The Licensee Failed to Control Protected Area Access Badges. and The Badges Were Taken Outside The Protected Area (Section 58.1)
50-369.370/98-06-01 IFI Root Cause and Corrective Actions for Failure of The 1A EDG Number 6R Cylinder Exhaust Valve Seat (Section M2.1)
50-369.370/98-02 URI Refueling Water Storage Tank Interior Coating Inspection (Section E4.1)
l CLOSED 50-369/97-03 LER Inoperability of Both Trains of the Control Room Ventilation System Due to Isolated Air Intake Valves (Section 08.1)
50-369.370/97 04 LER Main Steam Safety Valve Technical Specifications Inaccuracies (Section 08.2)
50-369/97-07 LER Unit 1 Reactor Trip During MODE 3 Power Range Instrumentation Calibration (Section M8.1)
50 369/97-0 LER Inoperability of the Auxiliary Feed Water (AFW)
Revision 0 & 1 System Due to Potential Air Entrainment (Section 08.3)
50-369.370/97-20-03 IFI Emergency Procedure Adequacy for Coping with a Tube Rupture In The RCP Thermal Barrier Heat Exchanger (Section 08.4)
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50-369/97-08-02 VIO Inadequate Procedure For Performing ACOT Testing (Section M8.1)
50-369,370/96-06-02 UR Refueling Practices Involving Full Core Off Load Into SFP (Section E8.1)-50-369,370/97-13-02 URI The Licensee Failed to Deactivate and/or Deny l Protected Area Access to Terminated Employees 1 (Section SB.1)
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50-369.370/97-13-03 URI The Licensee Failed to Control Protected Area Access Badges. In That The Badge Were Taken ,
Outside The Protected Area Unescorted By Security Personnel (Section SP 1) )
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l LIST OF ACRONYMS USED ACOT - Analog Channel Operational Test
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ALARA - As Low As Reasonably Achievable AFW -
Auxiliary Feed Water-AP- -
Abnormal Procedure ASME - American Society of Mechanical Engineers CAS -
Central' Alarm Station ,
CCTV - Closed Circuit Television CCW -
Component Cooling Water CFR -
Code of Federal Regulations CR -
Control Room DES -
Duke Engineering Services DRPI -
Digital Rod Position Indication DRWM - Dynamic Rod Worth Measurement EDG -
Emergency Diesel Generator EHRA - Extra High Radiation Aree ENS -
Emergency Notification System EP -
Emergency Procedure ES Engineered Safety Feature F' -
Fahrenheit FCF -
Framatome Cogema Fuels !
FMEA - Failure Modes and Effects Analysis GL -
Generic Letter GPM -
Gallons Per Minute HRA -
High Radiation Area IFI -
Inspector Followup Item IN -
Information Notice IR -
Inspection Report
ISLOCA - Interfacing System Loss of Coolant Accident KV -
Kilo-volt LER -
Licensee Event Report LLEA - Local Law Enforcement Agencies LOCA - Loss of Coolant Accident MOV -
Motor-0perated Vahe
< -MSSV - Main Steam Safety Valve
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MSIV - Main Steam Isolation Valve NCV -
Non-Cited Violation NRC
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-Nuclear Regulatory Commission NRR -
NRC Office of Nuclear Reactor Regulation NSD -
Nuclear Site Directive NSRB - Nuclear-Safety Review Board MUMARC - Nuclear Management and Resources Council OAC -
Operator Aid Computer OMP Operations Management Procedures PAP -
Personnel Access Portal PORC - Plant Operating Review Committee PIP -
Problem Investigation Process PM -
Preventive Maintenance PSP -
Duke Power Company Nuclear Security and Contingency Plan PT -
Periodic Testing OMSS - Qualified Maintenance Support Specialist RCA -
Radiogically Controlled Area RCS -
Reactor Coolant Pump
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RP -
Radiation Protection RWP -
Radiation Work Permit RWST - Refueling Water Storage Tank SAS -
Secondary Alarm Station SFP -
Spent Fuel Pool SGRP - Steam Generator Replacement Project SPP -
Security Plan Procedures SRWP - S)ecial Radiation Work Permit TB -
T1ermal Barrier TEDE - Total Effective Dose Equivalent TS -
Technical Specifications UFSAR - Updated Final Safety Analysis URI -
Unresolved Item VHRA - Very High Radiation Area V -
Volt VIO -
Violation WAPR - Workaround Problem Resolution WO -
Work Order ZPPT -
Zero Power Physics Testing
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