IR 05000498/1989030
| ML20248C683 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/26/1989 |
| From: | Holler E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20248C674 | List: |
| References | |
| 50-498-89-30, 50-499-89-30, NUDOCS 8910030544 | |
| Download: ML20248C683 (25) | |
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l APPENDIX U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-498/89-30 Operating License: NPF-76 50-499/89-30 NPF-80 Dockets: 50-498 l
50-499 Licensee: Houston Lighting & Power Company (HL&P)
P.O. Box 1700 Houston, Texas 77001 Facility Name: South Texas Project (STP), Units 1 and 2 Inspection At: STP, Matagorda County, Texas Inspection Conducted: August 1-31, 1989 Inspectors: J. E. Bess, Senior Resident Inspector, Unit 1 Project Section D, Division of Reactor Projects ,
l J. I. Tapia, Senior Resident Inspector, Unit 2 Project Section D, Division of Reactor Projects R. J. Evans, Resident Inspector, Unit 1, Project Section D Division of Reactor Projects D. L. Garrison, P.esident Inspector, Unit 2, Project Section D j Division of Reactor Projects i D. M. Hunnicutt, Senior Project Engineer, Project Section D Division of Reactor Projects l Approved: ,
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E. J/ Holler, Chief, Project Section D Date Division of Reactor Projects Inspection Sumary Inspection Conducted Auoust 1-31, 1989 (Report 50-498/89-30; 50-499/89-30)
Areas Inspected: Routine, unannounced inspection of plant status, onsite followup of plant events, licensee action on previous inspection findings, j onsite followup of written reports of nonroutine events, preparation for '
refueling, refueling activities, operational safety verification, monthly
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l -2-maintenance observations, monthly surveillance observations, and testing of piping support and restraint system Results: Within the areas inspected, no violations or deviations were identified. The licensee's response and followup to two Unit 2 trips and to a Unit 1 radiation streaming problem were complete and adequate in identifying root causes (paragraph 3). Preparation for Unit I refueling activities and conduct of the refueling _ activities were undertaken in.a safe and methodical manner (paragraphs 6 and 7). General housekeeping continues to be an area requiring attention (paragraph 8). bbintenance and surveillance activities were competently conducted (paragraphs 9 and 10). Several procedure discrepancies were identiffed to the licensee for implementation into the licensee's ongoing procedure upgrade program (paragraph 9). The licensee had acceptable procedures in place and conducted adequate testing of dynamic piping supports (paragraph 11).
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DETAILS I Persons Contacted
- A. K. Khusla, Senior Licensing Engineer
- D. N. Brown, Manager, SCD
- M. R. Wisenburg, General Manager, NSRB
- J. E. Geiger, General Manager, Nuclear Assurance
- J. T. Westermeir, General Manager, Adm. Sp *M. A. McBurnett, Licensing Manager
- W. H. Kinsey, Plant Manager
- V. A. Simonis, Plant Ops., Support Manager
- C A.- Ayala, Supervisor Licensing Engineeri
- J. Jump, Maintenance Manager
- T. J. Jordan, Plant Engineer, Manager
- D._ A. Learar, Reactor Support Manager
- S. L. Rosen Vice President, Nuclear Engineer and Construction
- R. W. Chewning, Vice President, Nuclear Operations
- H. W. Dannhardt, Lead Ops Spe *S. M. Dew, Manager, NPMM
- A. C. McIntyre, Manager, Support
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- J. R. Lovell Technical Services Manage *G. L. Parkey, IPS Manager-
- J. W. Loesch, Plant Operations Manager l In addition to the above, the NRC inspectors also held discussions with various licensee, architect engineer (AE), maintenance, and other contractor personnel during this inspectio * Denotes those individuals attending the exit interview conducted on 1 September 1,198 j Plant Status Unit 1 began this inspection period limited to 78 percent reactor thermal power because of feedwater pump operability problems. On August 4, 1989, Unit I began a controlled shutdown for a planned 55-day refueling outag ;
This is the first refueling outage for Unit 1. On August 5, 1989 Unit 1 i entered Mode 3. Mode 6 was entered on August 19, 1989. The licensee i planned to return Unit 1 to operation on September 28, 198 j l
Unit 2 began this inspection period at 65 percent reactor thermal power, limited by the maximum output current of the one functioning main step-up '
transformer. One of two parallel McGraw-Edison 700/784 MVA transformers l suffered an internal fault on July 13, 1989. As a result of the Unit l'
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refueling outage, one transformer from Unit I was moved over to Unit 2 to i
[ replace the damaged Unit 2 transformer. Unit 2 returned to 100 percent l reactor thermal power on August'21, 1989. On August 23, 1989, Unit 2
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tripped on low-low S/G 1evel, caused by the "C" feedwater isolation valve j l
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-4-going fully closed during a valve operability test. On August 25, 1989, Unit 2 returned to 100 percent reactor thermal power. On August 29, 1989, Unit 2 was manually tripped when operators responded to the loss of all three main feed pumps on an electrical overspeed signal. Unit 2 was returned to power operation on August 30, 1989. At the close of this inspection period, Unit 2 was at 99 percent reactor ther:nal powe . Onsite Followup of Plant Events (93702)
On August 14, 1989, during a routine radiological survey of the radiation controlled area (RCA), within the Unit 1 mechanical auxiliary building, the licensee found contamination in a stairwell. The stairwell is near noncontaminated auxiliary piping associated with the radwaste evaporato The apparent cause of the contamination was a valve lineup which allowed liquid radwaste to reach the inorganic basin located outside the power block. Details of this incident will be reported in NRC Inspection Report 50-498/89-3 On August 17, 1989, in response to an alarm for the Unit 1 turbine generator building (TGB) Sump No.1, a sample was collected and radioactive cobalt (Co-58) was detected at an activity level of 9.5E-08 uci/ml. The licensee collected two additional samples to verify the activity and that the chemistry high purity germanium detectors were not contaminated. The licensee took a sample on the oily waste surge tank (the TGB sumps discharge into this tank). An activity level of 2.3E-08 uci/ml Co-58 was indicate The licensee determined that the source of contamination was from the liquid waste (WL) discharging into the Unit 1 open loop cooling system (0C). A design error allowed processed liquid waste to discharge into a common header, a Further calculations by the licensee determined that the activity in the ;
OC system during a liquid waste discharge was high enough to cause detectable activity in the TGB No. 1 Somp. Valve 1-0C-283, the supply to DC chlorine analyzer, was found open and discharging into the TGB Sump No. 1. The valve was shut and tagged to prevent further discharging :
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into the sump. TGB Sump No. 2 was also sampled and no activity was foun The licensee conducted further tests during the next liquid waste release and confirmed that the source of contamination was the chlorine analyzer drain. A composite sample wa Let up at the oily waste (0W) discharge had been point and all release The samples collected licensee indicated will issue Stationno Co-58 Report Probleni activity (SPR) 85-0615 to further identify resolutions and corrective actions to this concer The inspector interviewed site personnel, reviewed procedures, and reviewed system drawings to ascertain if the liquid waste and OC were being operated correctly. No operator error, procedural violation, or system malfunctions were identified. Activity levels in the affected
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systems were below reportable limits. The inspectors will continue to monitor the licensee actions regarding radioactive and nonradioactive .j interface j On August 25,'1989, while performing radiation shield verification surveys during the transfer of spent fuel to the Unit 1 FHB, high radiation levels were detected on the exterior of the FHB. Contact dose rates of up to 15 R/hr were detected. The problem was detected during followup surveys performed after the required base points were monitored with the passage of the first assenbl The licensee determined the cause of this event to be penetrations in the exterior wall of the FHB which were not designed to account for radiation streaming associated with the fuel transfer- tube. -These penetrations are nearly in line with the transfer tube and were not provided with adequate shielding to prevent high dose rates outside the FHB. The initial design of the radiation shielding verification program did not identify these penetration The licensee determined that a whole body dose accumulation of up to 24 mrem could have resulted during a fuel bundle transfe The licensee perfomed a conservative dose calculation which assumed a person remained within 18 inches of. the source for 16 fuel transfers (approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />). The maximum whole body dose calculated was less than 400 mrem or about 80 percent of the annual whole body dose limit for members of the general publi The licensee tc.ok the following immediate actions after confirming the potential dose rates:
The Health Physics Division Manager notified station management and all fuel transfers were stoppe *
Plant layout and penetration drawings were reviewed by ar.gineering to identify other possible shielding deficiencies. A physical walkdown was completed by health physics management and engineering to verify these review *
An additional bundle was transferred to the FHB while radiation surveys were performed in areas identified in the drawing review and walkdown as having the. potential for inadequate shieldin * A tubelock barricade was erected around the penetrations and a security guard was posted to' limit access into the high radiation are *
Temporary shielding was installed to reduc,e the streaming dose rates foun " The tendon access areas and the roof of the FHB were conservatively posted as radiation areas as a result of the surveys performe _ _ . - - --_
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TLDs were exposed and processed in an effort to bound the possible doses that unmonitored personnel might have received as a result of-these penetration *
A limit of four bundles per hour was established during fuel :
transfers to limit highest possible dose accumulation rates to within l
=4 10 CFR 20 limits in the FHB and other area Health Physics continued surveys during fuel movement. At the end of this inspection period, no personnel had been identified as being overexposed during defueling of the reactor vessel. -The licensee is evaluating ,
engineering modifications to provide proper shielding. The inspectors ,
will continue to monitor the licensee response to this event. . This will
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beanopenitem(498;499/8930-01) pending inspector review of the final '
resolution of this matte On August 20, 1989, approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after Unit 1 exited mid-loop operation, a reactor plant operator reported smoke coming from Residual Heat Removal (RHR) Pump 'B." When the smoke was discovered, Train "B" was in service providing core cooling, Train "C" was in standby and Train "A" RHR was tagged out and drained in preparation for maintenance on the pump flange. Within 1 minute of the initial smoke report. Train "B" RHR pump was stopped and Train "C" RHR pump was started to provide core coolin With both RHR Pumps "A" and "B" inoperable, the plant entered TS 3.9. which requires immediate initiation of action to return two independent RHR loops to operability or to establish greater' than 23 feet of water above the reactor vessel flange. Immediate actions were initiated to restore Train "A" RHR loop to service. Approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after Train "B" was declared inoperable. Train "A" was declare operable and Action Statement 3.9.8.2 was exite The cause of the smoke on RHR Pump "B" appeared to have been caused by the ;
lower motor bearing overheating and causing the lubricating oil to catch !
fire. ThelicenseegeneratedStationProblemReport(SPR) 890620 to j investigate the cause of the lower motor bearing failure. The inspectors ;
will report the licensee's followup actions in a subsequent inspection repor !
On August 23, 1989, at 1:19 a.m., Unit 2 tripped from 100 percent reactor power. The trip occurred due to low-low water level in the "C" steam generator. This was caused by the "C" main feedwater isolation valve going fully closed instead of stopping at 90 percent open during the perfomance of a feedwater system valve operability test per Station
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l l Procedure 2 PSP 03-FW-0001, Revision As a result of the trip, auxiliary-
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feedwater was actuated and main steam was manually isolated to limit cooldown. The system functioned as expected after the trip and the unit :
Was taken to Mode 3 (hot standby) conditions without difficultie j i
I Subsequent investigations determined that the cause of the trip was a malfunction in the isolation valve test circuit. Specifically, the j 90 percent limit switch was not activated because the activating arms that ;
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-7- l are at tached to the valve stem by means of a collar device had moved out of position. The licensee initiated a reevaluation of the valve stem collar design. Regardless of the limit switch not worung, the licensee determined that the test' circuitry should have caused the valve to go open when the test button was released. Because this did not happen, the licensee initiated a modification to the test circuitry which will close j a contact after 1 second of depressing the test button and cause the valve j to go open. This timed contact is wired in parellel with the limit switch contact to make sure that the valve will go open if the limit switch fails to function. The 1 second of valve motion would equate to approximately l 25 percent closure. This amount of closure should not affect steam generator level since flow restriction does not begin to occur until the valve is at least 60 percent shu On August 29,1989, at 2 p.m., Unit 2 was manually tripped from 100 percent power when the reactor operators received feedwater flow / steam flow mismatch alarms on all four steam generators. Indications were also received of decreasing speeds on all three main feedwater pumps and of all three main feedwater pumps having received overspeed trips. The inspector observed the recovery efforts immediately after the trip and noted that the activities were handled professionally, and that the operations crew responded properly to the trip in accordance with Emergency Operating Procedure 2 POP 05-E0-E000, Revision 1, " Reactor Trip or Safety Injection."
The licensee subsequently determined that the cause of the simultaneous overspeed trip of all three main feedwater pumps was due to a momentary loss of 120 VAC power to the distribution panel (DP-048A) which supplies the control power circuit for the pumps. This loss of power resulted in an indicated, but not actual, electrical overspeed condition on all three pumps because the electrical overspeed protection circuitry is a
"de-energize to actuate" circuit. The electrical overspeed protection circuitry was a modification added to all Unit 1 and Unit 2 main feedwater pumps after the failure of a turbine-steam supply stop to shut destroyed .
the Unit 1 No.11 main feedwater pump. The original intent of the !
modification was to trip all feedwater pumps on a loss of power because a loss of power would also be sensed by the electrohydraulic control (EHC) !
controllers that operate the steam supply valve l l
With respect to the momentary loss of power, the licensee determined that the inverter supplying power to DP-048A malfunctioned. Two electronic !
logic cards (the static transfer logic card and the DC sensing logic card) l were found not to be functioning correctl As a result of the events on August 29, 1989, the licensee determined that ,
the electrical overspeed protection circuitry should be an " energize to j actuate" circuit. The licensee has initiated a design modification which, <
when implemented, will close or energize a contact in the 125 VDC pump trip circuit upon reaching the overspeed trip setpoint. The modification !
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should preclude spurious trips resulting from momentary losses of AC power to the distribution panel. The inspector will continue to monitor i implementation of this design chang i
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1 1-8-No violations or deviations were identified in this area of the inspectio . Licensee Action on Previous Inspection Findings (92701)
(Closed)OpenItem 498;499/8764-03: Component Cooling Water Heat Exchanger Capacity Testing The licensee committed to test component cooling water (CCW) heat exchangers when the plant was at normal operating temperature and pressure and evaluate the test results prior to exceeding 30 percent reactor powe The licensee completed the committed testing, recorded the test data, and evaluated the data and test results for the CCW heat exchangers heat load baseline performance test. The data and calculated results were reviewed by the inspector. The results of the test and calculated results were within acceptance criteria limits. Additional review of the overall 30 percent power plateau has been completed and was documented in Inspection Report 50-498/89-28; 50-499/89-28. This item is close (Closed)OpenItem 498/8775-04: Class IE Battery Periodic Service Tests Battery trains were tested using incorrect values because amperage values were incorrectly entered into Surveillance Test Procedure 1 PSP 06-DJ-0004,
"125 Volt Class 1E Battery Service Surveillance Test," for battery Trains "B," "C," and "D." TS 4.8.2.1.d and 4.8.2.2 require a 2-hour surveillance load profile test to be performed on engineered safety feature (ESF) batteries at least once during each 18-month perio Procedure 1 PSP 06-DJ-0004 was the controlling document. This procedure did not establish chat the capacity of each battery bank was adequate to supply and maintain the ESF loads for the " design duty cycle" as required by T The acceptance criteria for these 2-hour service tests were obtained from licensee engineering Letter ST-YB-HS-215. Load profile values were determined by Bechtel Engineering Calculation EC-5008 (EC-5008). The load profile values were based on EC-5008, Revision 3, and revised minimum acceptance battery voltage per Final Safety Analysis Report (FSAR) Change ,
Request 937. EC-5008, Revision 7 provided the correct amperage load j profile values for the battery load. The latest Bechtel Engineering Calculation (BEC) related to load profile value was EC-5008, Revision The EC-5008 calculations indicated that the battery was adequate to supply the referenced loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and maintain a minimum voltage of 106 volts direct current (VDC), or 1.8 VDC per cell. Final verification of adequacy was provided by the vendor, GNB Industrial Battery Company (GNB), in Document 4109-00008-CG A licensee investigation included verification that the service test i values used during startup preoperational testing were iiwe conservative !
than those in the later revisions of the BEC EC-5008. In accordance with j
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-9 the more conservative values, surveillance credit packages were prepare The surveillance credit packages were based on testing performed during Preoperational Tests 1-DJ-P-01, 1-DJ-p-02, 1-DJ-P-03, and 1-DJ-P-0 The licensee determined that the root cause of the error in the surveillance procedure, IPSP06-DJ-0004, was that the procedure was not reviewed and the load profile values changed to agree with EC-5008. The failure to have agreement between the test procedure and the calculated load profile / values wos caused by lack of attention to detail during the procedure preparation and review proces The licensee's corrective actions included the following:
A formal, documented procedure walkthrough program by the department which prepares a procedur A procedural adequacy review by the surveillance program coordinato *
An independent review by the quality assurance department and a TS verses surveillance procedure setpoint revie *
Listing as references, the engineering design documents which are pertinent to the performance of a plant surveillance procedur These references are included in the Nuclear Power Operations Department document reference tracking syste The licensee performed an engineering evaluation to determine if the use of an incorrect load profile could have had any deleterious effects on the batteries. Battery E1D11 was the only bank which was tested with load profile values significantly higher than the design requirements. The licensee's engineering evaluation detennined that the use of the incorrect load profile values did not have deleterious effects on Battery E1D11 or other ESF batterie The inspector reviewed the acceptance criteria, licensee's investigation report, corrective actions, and engineering evaluation. This item is close . Onsite followup of Written Report of Nonroutine Events at Power Reactor Facilities (92/00)
l (Closed)LER88-56: Failure to Test Containment Penetration Conductor Overcurrent Protective Devices On September 30, 1988, Unit I was in Mode 5 for a maintenance outage. The licensee was performing a review of containment penetration protection calculations. This review determined that some primary penetration protection circuit breakers had not been included in the surveillance test program as required by TS 4.8.4.1. Surveillance of electrical penetration protection breakers was developed from calculations which determined which
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-10-circuits had acequate overcurrent protection and which required periodic circuit breaker surveillance. Shortly before issuance of the Unit 1-Operating License (OL), the TS table identifying each breaker to be tested was removed from TS and placed in the FSAR as Table 8.3-1 FSAR Table 8.3-14 did not receive a prelicense verification; therefore, some errors _.were not identified. The failure to identify the errors resulted in the' omission of the breakers from the surveillance program. The
'icensee determined that 11 containment penetration conductor overcurrent protective devices had not been tested since receipt of the O The inspector reviewed the licensee's corrective actions, which included the following:
Surveillance credit packages were prepared or surveillance testing was completed on the penetration breakers which were not identified in FSAR Table 8.3-1 *
The licensee reviewed other tabular design data which was removed from the TS and placed in the FSAR prior to receipt of the OL to ensure that related surveillance reference the correct design document *
The calculation (Calculation No. 87-PH-001, " Electrical-Penetration Test Points") of penetration overcurrent protective device acceptance -
criteria was revised (Revision 9, dated December 30,1988) to include breakers identified as a result of this event. The calculation was applicable to Units 1 and *
The design engineering calculation on penetration protection was revised to include a note which required notifying the plant engineering department of any revisions to the calculatio *
The licensee established a Surveillance Program Task Force. This task force reviews the surveillance program and provides management with appropriate information. -
This item is close . Preparation for Refueling (60705)
A preparation for refueling inspection was performed to; ascertain the adequacy of licensee procedures for the conduct of refueling operations, administration requirements for control of refueling operations and plant conditions, and implementation of these controls. The inspector reviewed the following procedures in the areas listed to determine the licensee's readiness for refueling operations of Unit 1: Receipt, Inspection, and Storage of New Fuel
Procedure OPEP02-ZM-0002 "New Fuel Receipt, Inspection, and Storage," Revision The purpose of this procedure is to
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define the authorities and responsibilities of ' personnel involved in the receipt, inspection, and storage of new fuel at STP; to provide instructions for receiving, inspecting, and ]
storing fuel, and for completing of nuclear material transaction 1 reports (DOE /NRC Form 741); and to fulfill the requirements for j new fuel receipt, inspection, and storag Procedure IP0PO4-FH-0003, " Accident Involving New Fuel in Fuel 3 Handing Building," Revision This procedure provides .
guidelines for actions to be taken in the event that a new fuel .
assembly is damaged or dropped during fuel handling operation b. Fuel Handling, Transfer, and Core Verification
Procedure OPOP08-FH-0009, " Core Refueling," Revision The purpose of this procedure is to specify the conditions which shall exist while reactor core refueling is being performed and the _ sequence of events during reactor core refuelin *
Procedure OPMP04-FH-0001, " Feel Transfer System Inspection,"
Revision 1. This procedure provides instructions for maintenance and inspection of the fuel transfer syste J
Procedure IPOP04-FH-0001, " Accident Involving Spent Fuel in the Reactor Containment Building," Revision 1. This procedure provides guidelines for actions to be taken in the event that an irradiated (spent) fuel assembly is damaged or dropped during ;
fuel handling operations in the reactor containmen i
Procedure IP0PO4-FH-0002, " Accident Involving Spent Fuel in the fuel Handling Building," Revision 1. This procedure provides guidelines for actions to be taken in the event that a spent fuel assembly is damaged or dropped during fuel handling ]
operations in the FH *
Procedure 1 PSP 03-ZG-0004, " Containment Ventilation Isolation Operability Test," Revision 1. This procedure provided instructions to verify that each of the required containment ventilation isolation (CVI) valves actuates to its isolation position on CVI test signal, manual initiation, and high radiation test signal. The procedure requirements satisfy TS requirements stated in.4.6.3.2.b. 4.9.4.b and 4. !
" Procedure OPEP02-ZM-0005, " Internal Transfer of Fuel Assemblies,"
Revision 0. This procedure defines the requirements and provides instructions for documenting the transfer of fuel assemblie This procedure applies to all fuel assembly transfers other than i reactor core fueling and refueling operatiun * Procedure OPGP03-ZL-0004, " Physical Inventory of Fuel Assemblies,"
Revision 2. This procedure describes the administrative l
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fuel and fuel transferred into and out of the reactor vessel i (core) upon completion of reactor core loading operations to i verify that the reactor core loading pattern is correct and to record and verify the specific locations of all fuel assemblies ;
in the spent fuel pool (SFP) and the new fuel vaul l l
c. Inspection of Fuel to be Reused; Fuel Assembly Inspection j
B&W Fuel Company Procedure F0-001, " Setup Check Out, Disassembly and Packing of the ECH0-330 System," Revision {
This contractor procedure describes the sequence of operations ]
required to setup, checkout, disassemble, and pack the {
ultrasonic detection system (UDS). The procedure recognizes i that the UDS equipment will be assembled and installed in a potentially contaminated work area. Provisions for health physics (HP) monituring for airborne radioactive materials and surface contamination are included in the procedure. The procedure requires that personnel read and sign that they understand this procedure and Procedure F0-002 (discussed in j paragraph below) prior to commencement of work activitie B&W Company Procedure FO-002 " Operating Instruction for the )
ECH0-330 System," Revision 4. This contractor procedure '
provides instruction for the fuel inspection team in the operation of the UDS. The principle of operation of the UDS system is that the attenuation of an induced signal is greatly increased when water is present on the inside surface of a fuel ;
rod. The magnitude of the received UDS signal can be evaluated !
to distinguish between leaking (wet) and watertight (dry) fuel rod Contract requirements between the licensee and the contractor ;
requires that licensee personnel be involved in the operation of plant equipment (other than the contractor's UDS equipment), quality assurance (QA) of contractor activities, HP coverage of licensee and contractor personnel and equipment, handling and movement of fuel j assemblies, and related refueling and maintenance activities. The I contractor will operate the UDS equipment and evaluate the data and l associated result i The lirensee evaluated the apparent advantages of the UDS over " fuel sipping" to determine the integrity of the cladding. The UDS can ,
also measure the straightness and check for changes in diameter l (swelling) of individual fuel rods in each spent fuel assembly removed from the reactor core. The licensee's evaluation indicated that the UDS examination method to determine fuel rod cladding )
integrity is equal or superior to the " fuel sipping" metho l i
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-13-The inspector reviewed other related procedures, including procedures - 4 used to perform evolutions which involve unusual plant conditions, significant radiation exposure risks, the potential 'for introduction of foreign materials and/or objects into the reactor vessel and/or )
vital reactor plant systems, and the need for well-defined and coordinated responsibilities among the licensee's departments and the ,
contractors involved in the overall refueling, maintenance, and {
related activities. The licensee's approved procedures contained j prerequisite.s for commencement, conduction, and completion of refueling activities, including:
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Prerefueling surveillance testing required by T *
Surveillance required by TS during the conduct of refueling I activitie *
Provisions for inspection of irradiated fuel rods cladding conditions and sampling for fuel bowing, distortion, swelling, surface defects and damage, and buildup of foreign materials on the fuel rods.and assemblie *
Provisions for maintaining proper decay heat removal, subcritical configuration, and identification and location of each fuel assembl Provisions for detailed ODS and visual inspection ;
Provisions for maintaining housekeeping in the containment, !
spent fuel pool (SFP), and new fuel storage areas and control of loose object Provisions for training and qualification of personnel involved in the refueling activities, including reference'to TS requirements and licensed operator requirement d. Periodic Monitoring of Spent Fuel Pool Cooling Parameters Procedure IPOP02-FC-0001, " Spent Fuel Pool Cooling and Cleanup System," Revision 5, was reviewed. This procedure provides instructions for cooling the SFP, the in-containment storage area, and the reactor cavity; makeup to the SFP and other systems; and operation of the spent fuel storage gates inflatable seal The SFP cooling and cleaning system was walked down to verify the system was properly lined up for transfer of irradiated fuel to SF ;
The system was compared to Operating Procedure IPOP02-FC-0001 valve !
and electrical lineup checklists, and the system piping and instrument diagrams (P& ids). The P& ids used included the spent fuel cooling and cleanup system Drawings 5R219F05028-1, Revision 12, and SR219F05029-1, Revision 1 .
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-14-Observations made during the walkdowns included:
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The SFP contained water with a boron concentration greater than or equal to 2500 ppm (TS 3.9.1. TS interpretation TSI-0018 and Pra rdure OPGP03-ZO-0012 requirements). The pool contained water with a boron concentration of 2554 to 2577 ppm.- The boron concentration was being monitored every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by the chemical analysis departmen The SFP water level was being maintained greater than 23 feet over the top of the fuel seated in the storage racks (TS 3.9.11.1 requirements). The as-found level was approximately 27 feet over the top of the fuel storage rack The SFP level was required to be checked each shift by operations per IPSP03-ZQ-0002, Revision 9. " Modes 1-4 Operator Logs," and IPSP03-ZO-0003, Revision 6. " Modes 5, 6 Operator Logs."
i The SFP temperature was below the FSAR required limit for maximum water temperatur FSAR Table 9.1-1, " Spent Fuel Pool Cooling and Cleanup System Design Parameters," indicates the normal maximum allowed pool temperature is 140"F. The pool temperature before core unloading was 90*F, and after core unload it was 110*F. The SFP temperature is also required to be checked each shift by operations, per IPSP03-ZQ-0002 and IPSP03-ZQ-000 To prevent reverse flow in an idle loop, Step 5.1.1 of the SFP cooling and cleanup Procedure IP0P02-FC-0001, provides instructions to shut the Heat Exchanger Outlet Velve 1FC-0012A(B) if only one loop is started. When one of two operating loops is to be secured Step 5.2.2. instructs the operator to close the Pump Discharge Valve 1FC-0010A/B. The next step. Step 5.2.3, instructs the operator to open both IFC-0010A/B if both loops are secure Due to the wording of Sections 5.1 and 5.2, the possibility existed in which Valve 1FC-0012A/B would have been shut per Step 5.1.1 and not reopened per Section 5.2. Unit 2 Procedure 2 POP 02-FC-001, Revision 1, is worded differently and provides for opening Valve 2FC-0012A/B. The licensee stated that the Unit 1 procedure would be revise The two SFP cooling and cleanup system P81Ds had minor errors on each one. On 5R219F05028-1, Valve 1FC-0052 was shown as a test vent but was actually labeled a drain valve, and Valve 1FC-0081 was a drain valve, but the "D" designator was missing from the P&ID. On 5R219F05029-1, Valve 1FC-0034 was shown as a normally open valve but was actually a locked shut valve, and Valve 1FC-0076 was a test connection valve, but the "TC" designator was missing from the P&I _ _ _ _ _ _ _ _ _ _ _ _
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Cther minor SFP cooling system discrepancies included: Heat ,
Exchanger Outlet Throttle Valves IFC-0011A and 1FC-00118 had packing leaks with no MWR tags attached, the seal on Yalve 1FC-00118 was broken, the seal on Valve 1FC-0016B was missing, and typographical errors were noted on Valve 1FC-0076, Breaker Panel DPB-435, and Breaker Panel DPC-435 nameplate Additionally, Valve 1FC-0047A valve position' pointer was not in alignment with the open-close mark * Other minor Procedure IPOP02-FC-0001 discrepancies included:
the wrong valve locations were given on the valve checklist +or-Valves 1FC-0104 and 1FC-0087, and incorrect names were given %-
Valves 1FC-0071, IFC-0081, IFC-0052 and Motor Contro Center (MCC) Cubicles E184, Cubicle D4, and MCC E1C4, Cubicle F2, in the checklist The inspectors determined that the technical adequacy of the above and related licensee approved procedures met the requirements for conducting refueling operation No violations or deviations were identified during this portion of the inspection. The discrepancies noted during the walkdown were identified to the licensee for appropriate actio . Refueling Activities (60710)
The purpose of this inspection was to ascertain whether refueling activities were being controlled and conducted as required by TS and approved procedure During this inspection, the inspectors monitored the licensee's activities pertaining to refueling activities. Prior to the licensee beginning mid-loop operations, the following procedures were reviewed:
OPSP03-ZO-0027 Revision 7, " Locked Valve Deviation Log"
OPOP01-ZQ-0030, Revision 7, " Maintenance of Plant Operation Logbook Operability Tracking Log"
" OPGP03-ZO-0035, Revision 0, " Reduced RCS Inventory Operations"
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" OPOP03-ZG-0009, Revision 2, "Mid-Loop Operation" i i
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Unit 1 entered mid-loop operations on August 18, 1989. After the-installation of the steam generator nozzle dams, Unit 1 exited mid-loop operations on August 19, 1989. During the time the plant was in mid-loop ,
operations, the inspectors verified that mid-loop instrumentation and i other vital equipment required to keep the core' covered performed as- i require i I
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-16-Because this was the first refueling outage for ifnit 1, first-of-a-kind operations such as mid-loop operations were closely monitored by the inspectors. The following items were observed:
Through random selections, the inspectors verified that valves and other components, which are required by procedure to be closed, were administratively tagged close *
Shift supervisors, the mid-loop coordinator, and other key personnel were observed and interviewed regarding their specific responsibilitie The inspectors attended several shift briefings to ascertain if potential problems were being addressed and if oncoming plant personnel were aware of the plant statu The inspectors noted that a dedicated, reliable communication line was established between the control room and the containment buildin Prior to the start of the refueling process, the inspectors reviewed Procedure OPOP08-FH-0009, Revision 4 " Core Refueling." The purpose of this procedure is to specify the conditions which should exist while core refueling takes place. The following prerequisites were verified by the inspectors prior to core refueling:
Direct communication between the control room operator and personnel at the refueling stations in the reactor containment building (RCB)
and FHB had been established prior to core alteration '
Both source range neutron flux monitors were operable.-
Adequate lighting was provided to ensure clear visibility in the fuel ;
storage areas and reactor cor j
A core refueling status board for inventory of fuel assemblies during i refueling was available in the control roo A radiation work permit was issued for core refueling operation !
During refueling, the inspectors observed that boundaries and protective covering were in place to ensure that foreign objects would not fall into the open reactor vessel. Housekeeping conditions had been established in
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and around the pool areas in the RCB and FHB. Fuel was handled in-accordance with plant procedure The inspectors observed portions of the ultrasonic testing of fuel assemblies to identify if there were any rods which leaked. The testing was done by Babcock and Wilcox (B&W). Licensee personnel were observed to be involved in all phases of the ultrasonic testing of the fuel assemblie :
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-17-The ultrasonic test results indicated that there were two assemblies which had one rod in each assembly that leaked. The fuel assemblies identified as B-54 and A 10, will not be reloaded into the core. The licensee's original plan for core reload was to reload Assembly B-54 into the core and leave fuel Assembly A-10 in the spent fuel pit. Since both assemblies have an identified leaking rod, both fuel assemblies will remain in the spent fuel pi As part of the inspection, licensee compliance with TS Section 3.9,
" Refueling Operations," was verified. The parameters listed in Section 3.9 included providing water boron concentration limits, verifying that at least two source range monitors were operable, ensuring water levels were maintained above a certain level, and maintaining refueling support systems operable as necessary. The inspectors verified that Section 3.9 limits were met and were being monitored at the required intervals by the license During the review of TS Section 3.9 surveillance requirements, the inspector noted several items. P&ID 5R179F05007-1, Revision 14. " Chemical and Volume Control System," referenced operational Modes SA and SB in the notes section of the drawing. Operational nodes were defined in TS Table 1.2, but Table 1.2 did not list a SA or 5B as a mode (apparently the A and B refer to Mode 5 with the reactor coolant system loops either filled or unfilled). Also, Procedure Checklist IPOP02-CV-0005-2, Revision 5, "MAB CVCS Valve Alignment," listed Valve ICV-0198 position as open. A footnote on the position stated, " Locked closed during refueling i operations and Mode 5 with loops not filled." 1he valve was observed to be locked in place (throttled open). The as-found position of 1-CV-0198 l agreed with Drawing SR179F05007-1 requirements. The footnote on 1-CV-0198 i position in the checklist was apparently incorrec These apparent discrepancies were discussed with the license Prior to Unit 1 going into mid-loop operation, the inspector reviewed the !
following procedures: "0 POP 03-ZG-0009, Mid-Loop Opuration," Revision 0, and Nuclear Energy Services Procedure 80A9520, Revision 3, " Steam Generator Nozzle Dam Installation and Removal," and found both to be adequate for the installatio The steam generator dams (dams) are inflatable rubber devices in the shape of a pie dish with structural aluminum and stainless steel bracing. The
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devices are inserted during mid-loop operation to isolate the steam generators from the main coolant loops. This allows eddy current inspection of the steam generator tubes and refueling of the reactor at the same time. The uncrating, preliminary setup of the dams, and checkout of the pneumatic control modules for pressurizing the sealing bladders was found to be satisfactory. During removal of the steam generator manway covers 3 difficulty was encountered in that most of the bolts were seized; however, all of the manway covers were removed and the dams were installed. Reinstallation of the manway covers will be performed using studs and nuts in place of bolt __ _ _
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-18-The refueling activities.at the start of this outage were being performed !
in a safe and methodical manner. Precautions were being taken to ensure the safety of plant personne No violations or deviations were identified during this portion of the inspectio . (perational Safety Verification (71707)
The purpose of this inspection was to ensure that the facility was being operated safely and in conformance with license and regulatory requirements. This inspection also included verifying that selected activities of the licensee's radiological protection program were being implemented in conformance with requirements and procedures, and that the licensee was in compliance with its approved physical security pla The inspectors visited the control room on a daily basis when onsite and verified that control room staffing, operator behavior, shift turnover, ,
adherence to TS limiting conditions for operation, and overall control '
room decorum were being conducted in accordance with requirement During a tour of the Unit 2 control room, the inspector walked down the main control boards and reviewed all outstanding maintenance work q request (MWR) tags attached to the main control boards. The subject and l issue date of each MWR was noted. Subsequently, the inspectnr held J discussions with the work control center manager to assess the licensee's program for prioritizing MWR's. It was determined that a computerized main 3 control board report that statuses each MWR is generated on a weekly basis and was being reviewed by plant management to evaluate the effectiveness of MWR implementation. The inspector reviewed the latest issue of this report and found it to be an informative management tool that is providing a good control on the number and duration of MWR's that are associated .
with the main control boards. The following procedures were also reviewed i in conjunction with assessing the effectiveness of MWR implementation:
Station Procedure OPGP03-ZA-0090, Revision 0, " Work Process Program" !
Station Procedure OPGP03-ZA-0085, Revision 0, " Work Management System Procedure" Station Procedure OPGP03-ZA-0080, Revision 3, " Work Coordination Program"
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These procedures were found to provide effective instructions for establishing the process by which work activities are scheduled, coordinated, and tracke j I
Tours were conducted throughout various locations of the plant to observe work and operations in progress. Radiological work practices, posting of barriers, and proper use of personnel dosimetry were observed. Plant
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-19-walkdowns of Unit I were performed to ensure work practices, housekeeping, radiological posting, and other related activities were being properly controlled during the refueling outage. Items observed ~and reported to the licensee included:
During a routine tour of the Unit 1 FHB, the inspector noted water running out of the FHB truck bay door. Upon further investigation, the inspector found that from the inside of the building the door had a 1-inch gap between the bottom of the door and the grade sla Licensee personnel had placed a large fire hose across the area, thus the gap was not readily noticeable from the inside or the outside of the FHB truck bay. After reviewing the door design detail drawings, Richards - Wilcox Drawings 14926-4180-01001-AXX and 14926-4180-01002-AXX, the licensee determined that the exterior bottom seal either had not been installed or had subsequently been removed. Additionally, the dimension from the door bottom to the floor line, which was specified to be 3/8 inch, was actually 1 inc Since the building operates at a slightly negative interior pressure, a potential existed for fine dust and dirt to be drawn into the building. The missing bottom door seal was a safety Class 9 component, indicating that the seal was not a safety-related ite Nevertheless, the licensee issued Work Request No. XF-81507 to install a new seal. This work request has been completed and the j seal was replace Housekeeping was being maintained in most of the radiologically controlled areas of the plant. Some exceptions were observed: a q flammable liquid storage cabinet, containing'several cans of flammable liquids, was open and unattended in Room 067F (essential chiller room) of the mechanical auxiliary building (MAB); a 'second -
flammable liquid storage cabinet was open and unattended in the nonradioactive pipe chase Room 064 in the MAB; and housekeeping was not being maintained overall at the 10-foot elevation of the NAB and new fuel inspection Room 103 in the FHB. Miscellaneous trash, cables, hoses, rags, and tools were found scattered throughout the rooms.
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The following concerns in radiological controls were observed: a l
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radiation boundary sign and associated rope were laying on the floor with the sign face down in the letdown reheat heat exchanger room in l the MAB; inside the containment building, a " caution hi radiation
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area, fuel transfer, keep' out," sign and associated rope was laying- i on the floor near Emergency Sump-1C; and a protective clothing i laundry storage bin had its " caution-radioactive materials" sign !
inside the closed bin (personnel could not read the sign without-openingthebin). These discrepancies were identified to the licensee and immediately correcte ;
Procedure OPGP03-ZF-0004 Revision 1, " Control of Transient Fire -l Loads," Step 4.11 states, in part, combustible packing material, I including pallets and shipping containers, shall be removed from j l
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i-20-plant areas as soon as practical. A pallet was observed in the MA i near the hot tool room, a wood crate was observed in the RCB, and a wood crate was observed in'the truck bay of the FHB. The inspectors could not determine how long the combustible materials had been
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installe The licensee promptly. corrected the matter when informed by the inspecto Other observations included: standing water (more than' normal) was i observed in Rooms 206B (heating, ventilation, and air conditioning equipment room) of the electrical auxiliary building (EAB), i Room 064 of the MAB around-reactor containment chilled water pumps, and near the miscellaneous area supplementary air handling unit;in the MAB at the 60-foot elevation; a compressed gas cylinder was -
improperly tied off in the EAB electrical penetration area,10-foot elevation; two empty soft drink cans were found on, top of the EAB; and sand (remains of sandblasting work) was found to be clogging the' ,
storm drains on top of the FH The inspectors verified, on a sampling basis, that the licensee's security force was functioning in compliance with the approved physical security plan. Search equipment such as X-ray machines, metal detectors, and explosive detectors were observed to be operational. The inspectors noted that the protected area was well maintained and not compromised by erosion or unauthorized opening in the area barrie Even though the concerns identified by the inspectors were not significant safety concerns, the NRC inspectors- stressed to the licensee that housekeeping should be monitored more closel !
No violations or deviations were identified during this. portion of the inspectio :
9. Monthly Maintenance Observations (62703)
The inspectors reviewed and observed selected station maintenance activities ca safety-related systems and components to verify the maintenance was conducted in accordance with approved procedures, regulatory requirements, and the T The following activities were reviewed and observed:
l Low Head Safety Injection _ (LHSI) Pump "A" hot' leg recirculation flow i loop calibration, using Procedure OPMP08-SI-0927. "LHSI Pump A Hot i
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Leg _Reirc Flow Calibration (F-0927)," Revision 0, and Preventive Maintenance (PM)InstructionsIC-1-SI-86006087, Revision *
Qualified display processing system (QDPS) loop calibration, using Procedure OPMP08-ZI-0134, " Generic QDPS Loop Calibration,"
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Revision 1, and PM IC-1-EW-86002645, Revision l l
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Standby diesel generator fuel nozzle assemblies inspection, using Procedures OPSP04-DG-0001, " Standby Diesel Generator Inspection (During Shutdown)," Revision 3, and OPMP04-DG-0019, " Standby Diesel Generator. Fuel Nozzle Assembly Maintenance," Revision Generator stator cooling water pump rework, using' Work Request CC83208; Two instrument loop calibration checks were observed by the inspecto A generic QDPS loop calibration was performed on Flow Transmitter Al-EW-FT-6855, and a loop calibration check was perfomed on Flow Transmitter N1-SI-FT-0927. The inspector verified that the as-left data was within acceptance criteria limits, measuring and test equipment used was within their calibration cycles, and the acceptance criteria was in agreement with the licensee's instrument setpoint inde The following observations were noted by the inspectors and were discussed with the licensee:
Step 7.5.1 of OPMP08-SI-0927 instructed the technician to disconnect and label two field leads. The technician did not label the leads, as required by Step 7.5.1, because he felt the labelling was unnecessary. The configuration change log, Procedure Form OPGP03-ZH-0021-1, was used to document renovel and reinstallation of the field leads. The configuration change log clearly identified the field leads, making the labelling requirement unnecessar Step 7.9.1 of OPMP08-SI-0927 instructed the technician to remove all test equipment, then sign (double verification required) the calibration data package, indicating the performance of Step 7. Calibration Data Packages OPMP08-SI-0927-1 and -2 did not have the Step 7.9.1 signoffs. This forced the technicians to suspend the test and initiate a field change request to revise the procedure. The failure to include Step 7.9.1 signoffs in two Procedure OPMP08-SI-0927 data packages indicated the procedure was not closely reviewed prior to approva *
As part of the procedure review, the Scaling Manual Instrument Loop N1SI-F-0927, Revision 0, was compared to OPMP08-SI-0927 acceptance criteria. The tolerance for Computer Point SIFA0927 was noted to differ between the value listed in the procedure (247.49 gpm) and in the loop data- sheet (175 gpm) for the process variable value of 0.0 gpm. The tolerance listed in the procedure was less conservative than the loop data sheet tolerance. The actual value measured was 0.0 gpm, a value that was within both tolerance The licensee stated that Loop NISI-F-0927 will be revise * In Procedure OPMP08-ZI-0134, the calibration data package signoff steps were noted not to be in correct numerical order. This had no effect on the performance of the procedur . _ _ _ _ _ _ __
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-22-The Unit 1 No.12 diesel generator engine was scheduled for a limited overhaul to a certain extent during the refueling outage as planned preventative maintenance. The engine was inspected and work was performed to Procedure OPSP04-DG-0001, " Standby Diesel Generator Inspection (During Shutdown)." The procedure in Step 5.17 required the removal of all fuel nozzle assemblies, checking of opening (pop pressure)
and spray patterns, and reinstallation in accordance with Procedure OPMPO4-DG-0019, " Standby Diesel Generator Fuel Nozzle Assembly Maintenance."
The inspector witnessed the overhaul and setting of the injectors in the licensee maintenance shop. The settings on the injer+ars are accomplished by the stacking of shims which control spring pressure thus injector opening (3600 psi). The disassently, cleaning, and setting, as well as the
' equipment used, was in accordance with established procedure The Generator Stator Cooling Water Pump No.12 for Unit 2 was noted to be discolored around the bearing housing and coupling. Further inspection by the licensee noted that the bearing was running rough. On disassembly, the bearings were found to be galled and dirty. The coupling gear teeth were reduced to approximately one half the nomal size and the coupling grease was dry and was not lubricating the gea The inspector witnessed the disassembly and rebuilding of the unit. The bearing housing was the only salvageable part. The licensee used a coupling from Unit I which was out of service at the time. Work Order CC83208 and associated work instructions, work request, data sheets, and manufacturers' instructions were included in the package and the proper signoffs and signature points were satisfactory. The technicians performing the work were competent and knowledgeable, required administrative approvals were obtained, and the procedures were adequate to control the activitie No violations or deviations were identified during this portion of the inspectio . Monthly Surveillance Observations (61726)
Selected surveillance tests were observed to ascertain whether the activities were being conducted in accordance with TS and other requirements. The inspection included both direct observation of the tests and a procedure review to ensure that the tests were technically adequate. The inspections included:
- IPSP03-ZG-0004, Revision 1 " Containment Ventilation Isolation l
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l 1 PSP 06-DJ-0004, Revision 4. "125 Volt Class 1E Battery Service Surveillance Test"
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i-23-Specific items inspected during the tests included verifying that administrative approvals and tag-outs were obtained, reviewing test data for accuracy and completeness, and verifying that test instruments were within their current calibration cycle Items noted and reported to the licensee following observation of IPSP06-DJ-0004 included:
The note prior to Step 6.6 stated Steps 6.6 through 6.7.5 may be performed in any sequence. Step 6.7.2 instructed technicians to turn the electronic control / data logger unit (Battery Capacity Test Unit BCT-30) power on. Step 6.7.4 instructed the technician to l connect data logcer BCT-30 cables to the batteries. A note on the BCT-30 stated, " connect cables before applying power to BCT-30."
This suggested that the steps could not be performed in any sequence. The note prior to Step 6.6 required revision and Step 6.7.4 should have been performed before Step 6. *
The BCT-30 had four data logger cable connections that were )
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unlabeled. The connections should have been labeled (e.g. 1-30, 31-60, 61-90, and 91-120) to help prevent incorrect connection of the cable *
Step 6.7.4 provided instructions on connecting the data logger cobles I to the batteries. The instructions included connecting two leads called the " white number 00 lead" and " white lead with letters OA." i The BCT-30 had no such leads, but did have two white leads, one red l tipped and one black tipped. The two white leads were unlabele .
The step also stated, in part, to connect each of the numbered clip !
leads to its respectively numbered cell. .The step did not indicate whether the positive or negative terminals should t we been use Due to the unclear wording of Step 6.7.4, the technicians. hooked up the 59 battery leads on the wrong terminals. The technicians then reconnected the clips to the correct battery terminals when erroneous data was received at the BCT-3 *
The note prior to Step 6.10.4 stated, in part, that the BCT-30. test program would start over if the start pushbutton was pressed. The note referred to the automatic mode of operation. Step 6.1 provided instructions on how to manually perform the test. The note before Step 6.10.4 was not related to the step, therefore, it was in the wrong location in the procedur Items noted during the observation and review of IPSP03-ZG-0004 included:
Procedure 1 PSP 03-ZG-0004, Revision 1, was approved July 8, 1987. Per Procedure OPGP03-ZA-0010, Revision 11, " Plant Procedure Compliance, Implementation, and Review," biennial reviews were required on all safety-related procedures. Step 3.4.1 of OPGP03-ZA-0010 stated that quality-related procedures shall be reviewed on a periodic basis not >
to exceed 24 months in order to ensure that procedures are maintained current. Per discussions with the licensee Procedure IPSP03-ZG-0004, a safety-related procedure, did not have its biennial review
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b-24-completed at the time of its performance on August 23, 1989. The failure to perform the bier.nial review within 24 months was not considered a concern because: Procedure 1 PSP 03-ZG-0004 was in the process of being deleted (combined with other procedures into one test package); the procedure was technically correct and a station problem report (SPR) was written August 16, 1989, identifying several procedures that were overdue (including IPSP03-ZG-0004) on the biennial revie All personnel performing the tests were knowledgeable and competent, the tests were performed within required intervals, and a technical review of the procedures did not identify any additional concern No violations or deviations were identified during this portion of the inspectio . Testing Piping Support and Restraint Systems (70370)
An inspection was conducted on the licensee's ongoing inservice testing of dynamic piping supports (snubbers) in Unit 1. This inspection served to determine whether the surveillance of dynamic piping supports for safety-related piping was in accordance with facility TS Surveillance Requirement 4.7.9. Specifically, the sampling procedures were reviewed and functional testing was observed to verify that the test equipment was calibrated and that acceptable instructions and procedures were followe Per Letter ST-HL-AE-3130, dated June 23, 1989, the licensee notified the NRC of the sampling plan selected for testing. A 10 percent sample plan, as described in TS Section 4.7.9.e.1, was selected for each type of snubber installed. This plan called for at least 10 percent of the total of each type of snubber to be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of TS 4.7.9.f, an additional 10 percent of that type of snubber is required to be tested until no more failures are found or until all snubbers of that type have been functionally teste Conformance of the testing results with the requirements of this testing plan was verified during this inspection. The original sample of 38 snubbers contained 2 failures, one of each type. A second sample of .
38 snubbers was then tested and it contained 1 failure. A third sample of
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17 snubbers was tested ard no failures were identifie Functional bench testing on a Paul - Monroe Stationary Tert and Data Acquisition System, Model 4120, was witnessed on a Anchor Darling (AD)
5501 snubber from a 12-inch reactor coolant system line and on a AD 12501 snubber f rom a 10-inch feedwater line. All testing witnessed was performed in accordance with Station Procedure OPSP11-SN-0003, Revision 1,
" Mechanical Snubber Functional Test Procedure." It was verified that the l
force that initiated free movement of the snubber in both tension and compression was less than the specified maximum drag force, and that activation or restraining action was achieved within the specified range in both tension and compressio It was also verified that for the three
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1 C-25-l snubbers that failed to satisfy the functional test criteria, an engineering evaluation was performed to determine the cause of the failure and to determine if the components to which the inoperable snubbers were attached were adversely affecte No violations or deviations were identified during this portion of the inspectio . Exit Interview The NRC inspectors met with licensee representatives (denoted in paragraph 1) on Septenber 1,1989. The NRC inspectors summarized the scope and findings of the inspection. The licensee did not identify as proprietary any of the information provided to, or reviewed by, the NRC inspector I
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