IR 05000498/1988055
| ML20155E273 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/29/1988 |
| From: | Bess J, Garrison D, Tapia J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20155E271 | List: |
| References | |
| 50-498-88-55, 50-499-88-55, NUDOCS 8810120211 | |
| Download: ML20155E273 (27) | |
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APPENDIX :
U.S. NUCLEAR REGULATORY COMMISSION i REGION IV l
.NRC Inspection Report: 50-498/88-55 Operating License: NPF-76 !
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50-499/88-55 Construction Permit (CP): CPFR-129
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Dockets: 50-498 .
50-499 l
Licensee: Houston Lighting & Power Company (HL&P) {
P.O. Box 1700 l Houston, Texas 77001 ,
Facility Name: South Texas Project (STP), Units 1 and 2 f
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Inspection At: STP, Matagorda County, Texas Inspection Conducted: August 1-31, 1988 4 i
/// w ,'. # $d.7/88 Inspectors:[gJ.E.Bes~,SeniorResidentInspector, s ProjectDate/ I Section D Division of Reactor Projects 4( .
I bl M D. L. Garrison, Resident Inspector, Project T/h7bf8 Date/ !
Section 0, Division of Reactor Projects I
hl NW J. I. Tapia, Senior Resident Inspector, Project
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T 27 88 Date '
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, Sectior.D,DivisionofReactorProjects !
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- b k lam D. M. Hunnicutt, Senior Project Inspector, W27b98
[ Tate '
l l l Project Section D, Division of Reactor ,
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Projects ;
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Approved: /.
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E. Holler, Chief, Project Section D, Date [
DivisionofReactorProjects f I
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estotro211 881006 PDR ADOCK 05000498 t
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Inspection Summary Inspection Conducted August 1-31, 1988 (Report 50-498/88-551 Areas Inspected: Routine, unannounced inspection included plant status, licensee actions on previous inspection findings, review of licensee action on reported events (LERs), operational safety verification, F.SF system walkdown, maintenance observations, and surveillance observation Results: Althin the areas inspected, no violations or deviations were .
identi fie .
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Inspection Conducted August 1-31, 1988 (Report 50-499/88-55)
Areas Inspected: Routine, unannounced inspection included loose part monitoring system test, remote reactor shutdown test, control rod system test, 4 fuel handling building inspection, QA progran (test and experiments) review, design changes and modification, preoperational test witnessing, and allegation followu Results: Within the areas inspected, no violations or deviations were identifie '
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1 DETAILS Persons Contacted HL&P
- C. R. Beavers, Plant Engineer
- A. C. McIntyre, Manager, Support and Engineering
$5. Eldridge, Operations Support Manager
- J. Loesch, Plant Operations Manager
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- G. L. Parkey, Plant Superintendent, Unit 2
- J. A. Slabinski, Operations Quality Control (QC) Supervisor, Unit 2
- A. W. Harrison, Supervising Project Engineer
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- K. M. O'Gara, Project Compliance Engineer
- J. T. Westermeier, Project Manager
- S. L. Rosen Ge
- S.D.PhillIps,neralManagerOperationsSupport Licensing Engineer
- J. R. Lovell, Technical Service Manager
- S. M. Head, Supervisor, Licensing Engineer
- H. L. Duke, Staff Engineer
- D. R. '(eating, Quality Engineer Manager
- R. A. Gangluff, Chemical Analysis Supervisor
- T. E. Underwood, Chemical Operations Analysis Manager
- R. C. Hardison, Construction Supervisor
- L. Giles, Plant Operations Manager, Unit 2
- T. J. Jordan, Project Quility Assurance (QA) Manager
- S. M. Dew, Operations Support Manager
- G. Ondriska, Startup Supervisor
- H. F. Polfshak, Lead Engineer, Project Compliance Bechtel
- R. W. Miller, Project Quality Assurance Manager
- H. Herman, Quality Assurance Engineer
- C. R. O'Neil. Unit 2 Engineer Manager Ebasco
- R. A. Moore, Assistant Quality Control (QC) Site Supervisor
- E. P. Rosol, Slte Manager In addition to the above, the NRC inrpectors also held discussions with various licensee, architect engineer 4E), constructor and other
.:ontractor personnel during this inspectio * Denotes those individuals attendirig the exit interview conducted on September 2, 198 _ _ _ _ - _ _ _ __-______ - ___ ___ _ ___ _ - __ -
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2. Plant Status On August 3, 1988, at 11:30 a.m. (CDT), STP Unit 1, reached 100 percent reactor power and begin a 100-hour Nuclear Steam Supply Syste- (NSSS)
acceptance run. On August 6,1988, after more than 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> of continuous operation at 100 percent power, the speed controller for the No.12 Steam Generator Feed Pump Turbine (SGFPT) was lost due to a blown fuse. The loss of No. 12 SGFPT caused a feed flow / steam flow mismatch causing a turbine run back to approximately 87 percent reactor power before the plant was stabilized. This terminated the 100-hour NSSS acceptance ru On August 11, 1988, the plant entered Mode 3 (hot standby) to perform maintenance on the steam generator main feedwater regulating valves (MFRV). The maintenance included repacking the MFRVs and several
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other maintenance task On August 16, 1988, at approximately 10:54 a.m., Unit I reactor tripped from 100 percent reactor power approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> into the second attempt to complete the 100-hour NSSS acceptance run. The trip occurred during removal of a stator cooling pump from service to perform preventive
maintenance. The turbine tripped because of a low stator cooling
- differential pressure. The reactor trip was caused by the turbine tri The licensee will report this incident in greater detail pursuant to i 10 CFR 50.73. On August 24, 1988, STP, Unit 1, successfully completed
the 100-hour NSSS scceptance run and the licensee declared STP, Unit 1, in
, comercial operatio STP, Unit 2 is 98 percent complete. Hot functional Testing was cortpleted j on August 29, 1988. The first shipment of fuel assemblies was received on
September 8,1988. Preparations are currently in progress for the Structural Integrity and Integrated Leakage Rate Test . Licensee Action on Previous Inspection Findings (92701 and 92702)
The NRC inspector performed an onsite review of previous inspection findings to determine whether the licensee had taken approrriate corrective actions as stated in applicable licensee event rep.>rts (LERs).
Nvised procedures and plant logs. The NRC inspector also determined whether or not responses were adequate and met regulatory requirements, license conditions, and licensee comitment (Closed) Violation (498/8771-01 - EA 87-240): Safety Injection (SI) Cold Leg injection Valves Found Closed When Required ~to be Open The plant was operated in Mode 4 from Oct::her 31 to November 2,1987, (a total of 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br />) without two ope.able Emergency Core Cooling
. System (ECCS) flow paths as requirec by Technical Specification (TS) 3.4.3.1.c. This event was reported to the NRC by the licensee in LER 87-12. All three High Head SI cold leg injection valves were closed during this 51-hour period (TS 3.S.3.1 requires that with less than ',wo operable flow paths a minimum of two flow paths must be restored
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Hithin I hour or that the plant be in cold shutdown within the next l 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />). The licensee issued Field Change Requests (FCRs) and revised
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the following procedures:
. IPOP02-Ril-0001, "Residual Heat Removal System Operation,"
Revision 8, dated January 22, 1988 l
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. OPGP03-ZA-0063, "Plant Operations Shift Turnover," Revision 6, dated January 2., 1988 ;
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The NRC inspector reviewed the revised procedures and turnover log ;
changes. The procedure revisions and turnover log changes, when .
implemented, should prevent similar occurrences. The licensee completed '
appropriate remedial training for licensed operators. This violation is close (Closed) Violation (498/8806-01 - EA 87-240): Pressurizer low Pressure SI Setpoint Set Too Low Due to Procedural Error The plant entered Mode 3 on November 22, 1987, with all four pressurizer pressure-low trip channels set to trip at 1850 psi instead of at a minimum value of 1861 psi. This event was reported to the NRC by the licensee in LER 87-17. TE .'able 3.3-4, Item 1.e required the pressurizer pressure-low l setpoints to be set equal to or greater than 1869 psi with an allowable ;
value equal to or greater than 1861 psi. TS Tabic 3.3-3, Item ;
required the pressurizer pressure-low trip function to have a minimum of !
three safety injection trip channels operable prior to entering Mode 3 of ,
plant operation. The licensee issued FCRs and subsequently changed the I trip setpoint values from "1850 psi" to "1869 psi" and changed required TS :
allowable minimum and maximum values in the following piocedures:
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1 PSP 02-RC-0455, "Pressurizer Pressure Set 1 ACOT (P-0455)," l'
Revision 1 dated Septenter 14, 1987
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1 PSP 02-RC-0456, "Pressurizer Pressure Set 2 ACOT (P-0456) " L Revision 1, dated September 14, 1987
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IPSP02 RC-0457, "Pressurizer Pressure Set 3 ACOT (P-0457) " i Revision 1, dated September 15, 1987 [
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1 PSP 02-RC-0458, "Pressurizer Pressure Set 4 ACOT (P-0458),"
Revision 1. dated Septenter 15, 1987
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IPSP05-RC-0455, "Pressurizer Pressare Set 1 Calib'ation (P-0455) " i I
Revision 1, dated September 11, 1987 i IPSP05-RC-0456, "Pressurizer Pressure Set 2 Calibration (P-0456)," !
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Revision 1, dated September 19, 1987 (
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. IPSP05-RC-0457, ' Pressurizer Pressure Set 3 Calibration (P-0457),"
Revision 1, dated September 11, 1987
. IPSP05-RC-0458, "Pressurizer Pressure Set 4 Calibration (P-0418),"
Revision 1, dated September 11, 1987 l While reviewing and evaluating the pressurizer pressure-low trip setpoint values, the licensee discovered that the TS setpoint for power range flux high positive rate was not covered by surveillance procedures. The :
licensee issued FCRs (Nos. 88-1296,88-1297,88-1298,and88-1299)to '
revise the following value of 109 percent)to procedures correct this(reset highand omission rangeprovidetrip setpoints completion to of TS Regulatory Guide 1.68, "Initial Test Programs for Water Cooled Nuclear Power Plants," startup sequence:
. IPSP02-NI-0041, "Power Range Neutron Flux Channel * ACOT (N-0041)," l Revision 2, dated February 5,1988 ;
. IPSP02-N1-0042, "Power Range Neutron Flux Channel II ACOT (N-0042)," l Revision 2, dated February 5,1988
. 1 PSP 02-NI-0043, "Power Range Neutron Flux Channel III ACOT (N-0043),"
Revision 2, dated February 5,1988
. 1 PSP 02-NI-0044, "Power Range Neutron Flux Channel !Y ACOT (N-0044)," ;
Revision 2 dated February 5,1988 l r
The NRC inspector reviewed procedure changes, which corrected these l errors, and licensee audit rcports. These audit reports stated that no !
additional TS translation errors existed. The licensee changed the '
l program, subsequent to this event, to require verification of implementation of TS changes by the Nuclear Assurance Department. This t
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violation is close I (Closed) Violation (498/8804-01): Failure to Provide Adecuate [
Procedure for Operational Control of Plant System [
The licensee reviewed Station Procedure 1 POP 03-ZG 0003, "Secondary Plant Startup " and fouad it inadequate in that it did not specify temperature or pressure level in thelimitations for using(the steam generators SG). main feedwater Procedure system towas IPOP03-ZG-0003 increase revised water '
(Revision 5) by adding Step This step required maintaining the teaerator pressure less than 50 psig to ensure deaerator temperature :
remained less than SG temperature during plant heatups. Support Engineering set up a task force made up of Bechtel, Westinghouse, and Houston Lighting and Power (HL&P) representatives to review the secondary system plant operating procedures with respect to potential hydraulic '
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transients (water hamer). The coments from this task force were evaluated and incorporated, as applicable, into plant procedures and training requirements. The licensee conducted licensed operator training i
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7 in heat transfer and fluid flow related to water hammer. The training related to water hammer is docunented in the Course Attendance Station Requal Training Records. This violation is close (Closed) Violation (498/8809-04): Four Cases Where TS Surveillance Requirements Were Not Met
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The NRC inspector reviewed procedure changes, applicable LERs and related
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changir.g test frequency, and methods for approving changes to
! surveillanc(s. The following was noted regarding the four cases cited in :
the violation:
! Periodic surveillance test for Pressure Transmitter CV-PT-204 had not
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been develope A surveillance procedure had not been developed for CV-PT-204 because ,
the instrument had not been addressed in the STP TS. This event was i reported to the NRC in LER 88-10. The LER was based on the instrument loop being inoperable because it could not be calibrate The condition was not recognized during a review of the Maintenance Vork Request (MWR). Correctise action required an evaluation of the
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program for generation and approval of MWRs. The licensee revised Procedure OPGP03-ZM4003, "Meintenance Work Request Program,"
- (Revision 16). Step 4.12.3 of the procedure now requires the approval authority review of the work scope to ensure that the ,
associated equipment does not affect the TS and includet che: king !
! design documents as appliccbl :
1 Missed surveillance test on 118 Essential Chill Water Pum ,
This missed surveillance was caused by inadegaate control of the
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serveillence test package. Also, the missed surveillance test did i I not appe2r on the Overdue Report. This event was reported to the NRC l l in LER 88-31. Procedure OPGP03-ZE-0004 "Plant Surveillance i j Program," Revision 7, created the positions of Divisional i Surveillance Coordinators and assigned these coordinators the I
{ responsibility for following surveillance test packages assigned to ;
the respective divisions. The methodology for changing the test ,
! f requency and the trethod for approving changqs to surseillance 1 dates have been added te P'scedure OPGP03-ZA-0055, "Plant i i Sueveillance Scheduling," acvision ,
! Missing stroke tire in pump cnd valve inservice test (IST) plan.
l This problem was caused by failure to incorporate revisions to the i j IST plan into implementing proccoures. This event was reported to ;
the NRC in LER 88.12. A temporary procedure.1 TSP 03-CV-0001, l
> "Charging Flow Control Valve FCV-0205 Operability Test," Revision 0, l
) corrected the deficiency and allowed the valve to be declared ;
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operabl Revision 3 to this procedure incorporated the stroke time limiting valu Procedure 1 PSP 03-CV-001). "Chemicai and Volume '
Control System Valve Operability Test (Cold Shutdown)," Revision 4 included provisions for obtaining the open stroke time for CV-FCV-020 Proceduro OPGP03-ZE-0021, "Inservice Testing Frogram for Valves," Revision 3, included controls to review the implemerting procedur The licensee's reviews of the IST and other pump and valve inservice surveillance tests did r.ot identity any other discrepancies, No periodic surveillance +.esting of feedwater isolation / turbine trip logic channel time delay relay This event rewlted from personnel error and was reported tc the NRC in LER 88-13. Procedure OPGP03-ZA-0002 "nlant Procetnes," wu revised (Revision 11) to require a second, independent technical l review of procedures written or ravised to ensure accuracy and l
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adequacy. Procedure OPGD03-ZE-0005, " Plant Surveillance Procedure Preparution," was revised (Revision 9) to require that available infonnation be reviewed to ensure thit the circuit being tested by the procedure is physically the circuit required to be tes'.ed by the l
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surveillance requirement This violation is close (Closed) Violation (498/8809-05): Failure _ tr implement 15 Action Requ i remen,t s, l The NRC inspector reviewed procedure changes, applicable LER and related corractive actions, licensee evaluations and reviews, review of the operability requirenents for containment isolatior, valves with the Licensed Operators (documented in simulator scenarior and the related course attendance records) between March 28 and April H.1988. Also, the latest revision to the Licensed Operator Training Program was reviewed te ensure that TS are taught with emphasis on practical application The licensee failed to recognin and initiate a coo'idown to Hode 4 when two main steam isols,ttion valvt.s (HSIVs) were found to be inoperable (TS
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LC0 3.7.1.5 permits one MSIV to be inoperable in Mode 3). The MSIVs
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became inoperable when the valve packing glands on the two MSIVs were over
! tightened. The overtightened packing glands rendered both MSIVs inoperable while the plant was in Mode 3. This event was reported to the
'E in LER 88-15. Procedure OPOP01-ZQ-0030 "Maintenance of Plant Operations Logbooks," Revision 3. when implem nted, should prevent recurrences of similar event This violation is close (Closed) Violation (498/C824-04): A Plant Procedure was not Followed in
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that Bound 3 ries were not Ccerectly Cuntrolled Plant Procedure OPGP03-ZO-0001. "Cquipmnt Clearance," Revision 7 described the requirements for controlling system boundaries and the
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logging of components within these boundaries. The lubricating (lube oil)
oil system on No. 13 emergency diesel generator (EDG) was cleared but persennel did not properly execute the requirements of the procedure. The licensee failed to list the components inside the boundary valves. This resulted in an incomplete valve alignment cneck of the EDG 'ube oil system. When the licensee placed this system back in service,
< appro/imately 1000 gallons of lubricating oil was pumped fron the lube nil sump onto the floor of the diesel generator building.
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TM NRC inspector reviewed the licensee's directivt to operations perst.nnel reinforcing the need to tollow procedures and the policy for having aa observer in the vitinitj when starting or stopping major componentt.. Operations personnel signed and dated a nenorandum to
- ecknowledge rect.Qt and understanding of the "Equipment Clearance
Procedu re. " The licensee's corrective actions should prevent similar i occurrer:es. The licensee completed appropriate renedial training for oport.tions personnel. This violation is close (Closed) Open Item (499/88?1-02): Training of Construction Personnel in theUseofHoistingDevicesinUnitj i The NRC inspector reviewed the licensee's lesson plan, "Proper Use of Hoisting Devices." This plan (training supplement) provided infomation on proper use of certsia types of hoisting device.,, including electrical,
, hydraulic, and hand operated equipment used by various craf ts. The lesson plan included introduction to types of equipment, littirg requiremrats, pr:per storace of liiting equipment, and precautions required during
, instal:ation and operating equipnent. Appropriate licensee craft personnel were trained in the use of hoistino devices in accoraance with the lesson pla This open item is closed, ho violations or deviations were identified in the review of this .
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Inspection area.
i 3. _ Review of Licensee Action on Reported Esents (LERS) (9E/00]
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Tne NRC inspector perfomed onsite review on the following LERs to
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determine whether the licensee had taken oppropriato corrective tctions as
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stated in the LERs and whether responses to the events were adequate and q
met regulatory t equirements, license c:nditions, and licensee comitment (t,losed)LER87-10: Fuel 1:andling Ba11 ding (FHB) Raciation Monito Causes j , Engineering Safety, Feature (ESFf Acto:>fion On July 20, 1987, and again on September .?6, 1937, with Unit 1 In Mode 5 ,
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a THB Ventilation System auto-ectaation to filtration mode occurred due to
i tu apparent failurs (a "loss of counting ability") of FHB atmcsphere radiation monitor ( A1RA RT-c0?b).
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The cause of this spurious auto-actuation could not be identified. The
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plant computer (Pr) baards were inspected and the electrical contacts ;
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cleaned, the input / output and preamplifier PC boards were replaced and the calibratiun surveillance procedure was satisfactorily performed. This ,
monitor (RT-8035) has functioned properly since the "loss of counts" that .
occurred in September 1987. This LER is close [
l (Cic3ed)LER88-01: Reactor Coolant pumo f tart with Secondary Water -
1 Temperature Greater than 50 Degrees Fahrbnheit Above the Reactor >
j [oolant System (RCS), and Pressurizer FORV Actuation
On January 2,1988, with Unit 1 in Mode 5 and the RCS solid RCS Pump 1A was started wnile filling and venting the RC The RCS cold leg temperature was less than 350*F. The secondary water temperature was
. greater than 50*F above the RCS cold leg temperature. The temperature differenc2s betnen the RCS cold leg and the secondary system temperature
- exceeded the limits stated in TS Section 3.4.1.4.1. Starting RCS Pump 1A under these conditions resulted in the RCS pressure exceeding the Cold ,
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Overoressure Mitigation System pressure setpoint. The overpressure caused !
Pressurizer PORV PCV-0656A to open morentaril i The licensee revised Procedure IPOP02-RC-0003, "Filling and Venting the
Reactoi- Coolant System," Revision 5, dated January 4, 1988, to include a specific instruction on how to take RCS and secondary side water ,
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temperature measurements, and operators were trained on this even '
Operator requalification training included temperature stratification and the difference between SG vessel surface temperatures and bulk water temperature. Licensee Engineering evaluated the incident and determined that no RCS structural damage had occurred. The TS limits (Section 2.1.2,
"RCS pressure," and Section 2.2.2, "Pressure and Temperature Limits") for I
RCS pressure as a function of temperature (Appendix G of ASME B&PV Code,
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Section !!!) were not exceeded. This LER is closed,
t (Closed)LER88-05: Inadequate Surveillance Performed on a Control Room l Intake Air Radioactivity Monitor
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j On January 10, 1988, with Unit 1 in Mode 5, the licensee failed to maintain a surveillance interval of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as specified in TS , Section 4.3.1 for a control room intake air radioactivity monitor '
(RT-8034). Operators had recorded the Hi-Alarm setpoint rather than the :
i actual gaseous activity for the channel check for five shifts. The l monthly surveillance procedure did not ensure that the monitor display was returned to normal following a routine surveillance en Train "A" of the Control Room Emergency Ventilation System (CREVS).
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The licensee's analysis of this event detemined that the Control Room l
! Intake Air Radioactivity Monitors were operable. The ronitor would have ;
I I initiated an Engineered Safety Feature (ESF) actuation and placed the CREVS in the recirculation mode, if an actuation signal had been received, l !
- Ti.e Itcensee reviewed and revised operations procedures which affect the
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Radiation Monituring Panel displays to assure that upon completion of i
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tests and surveillances the monitors display the appropriate activity values. Operating log procedures were reviewed and revised to assure that ;
j radiation monitor readings are compared to limits and/or against each '
other. Operators received additional guidance on detailed operation of
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i the radiation monitors. This LER is close l
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(Closed)LER88-12: Failure to Fully Implement TS Surveillance
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Requirements Due to Procedural Deficiency
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On January 29 1988, with Unit 1 in Mode 4, prior to initial criticality,
- a limiting vaIue fnr stroke time (measurement and recording of the open strol e time) on Valve CV-FCV-0205 was omitte The Itcensee prepared a temporary procedure,1 TSP 03-CV-0001, "Charging
Flow Control Valve FCV-0205 Operability Test," and perfonned an 1 operability test of the valve with the stroke timing requirement ;
incorporated. The licensee revised Procedure OPGP03-ZE-0021. "Inservice i
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Testing Program for Valves," Revision 3 (to inelade stroke timing for
! Valve CV-FCV-0205), and Procedure OPGP03-CV-0011. "Chemical and Volume Control System Valve Operability Test (Cold Shutdown)," Revision 3 (to '
1 include additional review controls), to correct the deficient condition.
q The licensee perfonned independent reviews of other purrp and valve
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inservice surveillance tests. The reviews were performed to assure 1 agreement between the implementing procedures and the IST Pla In i addition, the stroke time limiting value for CV-FCV-0205 was incorporated
, into Revision 3 of the IST Plan. This LER is close L
- L (Closed)LER88-24: During Review of Solid State Protection aystem (SSPS) l
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Design the Lir.ensee Discovered that 51 can be Blocked tJnder Certain i
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On March 1.o.1988, with Unit 1 in Mode 3, a design error was discovere : The erro" could cause a blockage of the Si actuation on Train A B, or C i j when the safeguards test cabinet master reset switch was operated and the l reactor trip breakers were ope !
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! A design change had been issued to modify the circuitry. The Shift
- Supervisor (SS) ensured that the testing policy for the SSPS was carried j
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out on Unit 1. The design change had been incorporated and installation L of the design change verified. This LER is close .
! (Clostd) lek 88-26: Degraded Undervoltage Coincident with Si j Su ve llance D,uiciency Due to a Deficiynt Procedure [
On December 12, 1987, with Unit 1 in Mode 4, prior to initial criticality, ,
the licensee determined that the Trip Actuation Device Operational !
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. Test (TADOT) on degraded undervoltage coincident with 51 had not been '.
! tested (TS Table 4.3-2, item 8.b) as required. All three ESF busses were
- declared inoperabl l
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l The cause of the event was determined to be a deficient surveillance procedure resulting from a personnel error in interpreting the requirements of the monthly TADOT. A new procedure, IPSP06-PK0005,
"4.16KV Class 1E Tolerable Degraded Voltage Coincident With SI And Sustained Degraded Voltage Relay Channel Calibration /TADOT-Channel 1,"
Revision 1 dated July 14, 1988, was prepared and satisfactorily performed on each ESF bu The licensee established a station policy for relay testing to provide additional definitions and guidance for use by plant personnel in implementation of the TS surveillance requirements and reviews. The licensee conducted independent reviews of other instrumentation & Controir (!&C) and electrical surveillance tests and procedures to assure that other required testing requirements were incorporated in the surveillance procedures. This LER is close (Closed) LER 88-30: Toxic Gas Monitor High Alarm placed Control Room Envelope on Toxic Gas Recirculation On May 6, 1988, with Unit 1 in Mode 5, an automatic actuation of the control room ventilation to recirculation mode occurred as a result of a high level trip of the hydrochloric acid (hcl) channel on gas analyzer XE-932 The cause of the actuation could not be identified. Available evidence suggested that a "puff" of hcl gas or gaseous hydrocarbon was detected by
+he analyzer. The licensee verified that the toxic gas analyzers nasponded to hcl samples. Public address announcements of toxic gas actuations will be made. A memorandum emphasizing the sensitivity of the toxic gas analyzers to gasts and fumes and the need to notify the control room of activities that produce 'ases or fumes in or around the power blocx was issued. This LER is c u ed. These announcements will require personnel involved in a :tivities producing any gases or fumes to imediately contact the control roon (this requirement should assist plant personnel in determining the sources of gases that cause any future actuations).
No violations or deviations were identified in the review of this inspection are Operational fafety Verificacion - Unit 1 (71707)
The purpose of this inspection was to ensure that the facility is being operated in conformance with the requirements established under 10 CFR Part 50 and the TS. This inspection also included verifying t % t selected activities of the licensee's radiological protection programs were being inplemented in conformance with plant policies and procedures, and the licensee's compliance with the approved physical security pla The NRC inspector performed inspection in the control room on a daily basis and verified:
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. Proper control room staff was maintaine . Operators wera adhering to approved procedures for ongoing activitie t
. Operability of reactor protective systems and engineered safety components was as require . Control room was free from distractions such as nonwork-related reading material .
The NRC inspector toured various areas of the plant to observe work in i progress. Posting of Radiation Work Permits (RWPs), the proper ase of personnel dosimetry, and the correct methods for frisking when exiting the radiation protected area (RPA) were observe The NRC inspector verified that the licensee's security plan was being implemented in accordance with its security program. The NRC inspector observed that packages and personnel were properly checked prior to entry 1 into the protected area (PA), illumination in the PA was adequate to observe all areas during hours of darkness, and personnel inside the PA had proper identification badge No violations or deviations were identifie . Engineered Safety Fea,urt (ESF) System Walkdown - Unit 1 ('1710)
The NRC inspector conducted a walkdown of the accessible portions of Train "A" of the safety injection (SI) system to independently verify the operability of the system. A review was perfonned to confinn that the licensee's system operating procedure matched plant drawings and the ;
as-built configuration. Equipment condition, valve and breaker positions, r
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housekeeping, labeling, pennanent instrument indication and calibration, and operability of support systems essential to actuation of the ESF system were observe l The NRC inspector identified the following items to licensee managenent: ;
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The label on Breaker EIA2-P2 indicated as follows, "LHSI Pump 1A !
Recirc. to RWST M0V-0014A." The SI electrical lineup indicated this breaker supplied power to LHS! Pump 1A miniflow valv .
The label on Breaker EIA2-P1 indicated as follows, "LHS! Pump 1A Recirc. to RWST M0Y-0013A." The SI electrical lineup indicated this breaker supplied power to LHSI Purp 1A miniflow Valve MOV-0013 .
The label on Bre .er E1A2-Q3 indicated as follows, "HHS! Pump 1A ,
Recirc. to RWST NOV-0012A." The SI electrical lineup indicated this breakt.r supplied power to HHS! Pump 1A miniflow Valve MOV-0012A.
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The label on Breaker E1A2-02 indicated as follows, "HHSI Pump 1A Recire. to RWST MOV-0013A." The SI electrical lineup indicated this breaker supplied power to HHSI Pump 1A miniflow Valve MOV-0013 .
The label on Breaker E1A2-N2 indicated as follows, "Containment Sump Isolation Valve MOV-0016A." The SI electrical lineup indicated this breaker supplied power to Emergency Sump Suctio . Valve S10071A showed indication of mtnor leakage as evidenced by the crystallization of boron around the valve bod During previous inspections, the NRC inspector has identified these types of discrepancies to the licensee. In response to the concerns expressed by the NRC inspector, the licensee stated that a plan to walkdown all safety-related systems had been implemented. The licensee also stated that the criteria to be used when evaluating the systems are:
. Compare component / valve labeling with piping and instrument diagrams (P& ids).
. Compare breaker labeling with electrical wiring diagram . Compare "Noun Namer" on labeling and related procedure . Install labeling in the fiel . Make required procedural change The licensee estimated that the above items would be completed by November 30, 198 No violations or deviations were identifie . Monthly Maintenance Observation - Unit 1 (62703)
The station maintenance activities listed below were observed and documentation was reviewed to ascertain if the activities were conducted in accordance with approved procedure On August 12, 1988, Maintenance Work Requests (MWRs) FW-65273 FW-46608, and FW-46609 were initiated to perfonn maintenance on the four Main Feedwater Regulating Valves (MFRVs). The valves were leaking excessively during power operation. The NRC inspector observed the repacking of the MFRVs and verified that the work was being perfonned in act.ordance with Procedure, OPHP04-ZG-0003, "General Valve Repacking," Revision The NRC inspector concluded that the work packages provided adequate instructions to maintenance personnel for the circumstance No violations or deviations were identifie . Surseillance Observations - Unit 1 (61726)
The NRC inspector observed selected portions of the surveillances listed below to verify that the activities were being performed in accordance with the TS and surveillance procedures. The applicable procedures were
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reviewed for adequacy. Test instrumentation was verified to be in calibration, and test data was reviewed for accuracy and completenes The inspector verified that deficiencies identified were properly resolve . Procedure 1 PSP 03-EA-0002, "ESF Power Availability," Revision .
. Procedure 1 PSP 03-SI-0013. "Accumulator Isolation Valve Verification,"
Revision 1.
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The NRC inspector verified the following itens during the inspection:
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. Test results were reviewed by personnel other then the persons directing the tes ,
. The surveillance testing was conpleted at the required frequency per TS requirements, i . Testing was performed by qualified personnel using approved j procedures.
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No violations or deviations were identifie i
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8. Loose Parts Monitoring Systen Test Unit 2 (70450) l t
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The NRC inspector reviewed Acceptance Test Procedure 2-IB-A-01, "Loose
! Parts Monitoring System " Revision 0, dated June 16, 1988. The purpose of f
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+ this te;st was to verify that the Vibration and Loose Parts Monitor (V& LPM)
I would provide the information and initiate the alarins required to alert i l operations personnel of unusual occurrences within the .; cope of the '
- measuring devices in the V&LPH syste The test was conducted as required by the test procedure; Regulatory Guide 1.68 '-Initial Test Programs for <
Water Cooled Nuclear Power Plants"; Regulatory Guide 1.133, 1981, ;
Revision 1 "Loose Parts Detection Program for the Primary System of ,
j Light-Water-Cooled Reactors"; and Final Safety Analysis Report (FSAR),
- Section 4.4.6.4 - Amendment 6 t
The hRC inspectors observed portions of this V&LPH test and determined l
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that the test was pcrformed in accordance with the approved procedure.
! Regulatory Guides 1.133 and 1.68, applicable portions of the FSAR, and ;
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! proposed TS. The V&LPH system was adequately tested to ensure that NRC requirements and licensee comitments were satisfied. The test equipment was properly installed and calibrated. Approved procedures were available to the personnel conducting the test. The test data was collected and recorded in accordance with this procedure. The licensee personnel were qualified to conduct this test, record the test data, and l evaluate the test result ;
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The V&LPH system responded to test signals, operated in accordance with !
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established parameters, remained within calibration requirements, and t demonstrated the ability to monitor the various plant component !
, r The test results were reviewed and approved by qualified licensee !
i personnel. The NRC inspector's review of the approved test results agreed '
with the licensee's evaluation of this test.
. No violations or deviations were identified in the review of this !
inspection are [
. t 9. Remote Reactor Shutdown Test Unit 2 (70452) j I The NRC inspectors reviewed preoperational Test Procedure 2-RC-P-10, i
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"Remote Shutdown," Revision 1, dated February 5,1988. The purpose of !
this test was to demonstrate the ability to perform a cold shutdown from !
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outside the control room. The test was ccaducted in accordarce with this I i
test procedure; Regulatory Guide 1.68, "Initial Test Programs for Water I
Cooled Nuclear Power Plants"; FSA.R. Section 14.2.12.2(98),"RCSHot l'
l Functional Preoperational Test Sumary," Test Objective 21 and Method 19 (a through e); and FSAR, Section 7.4.1.9, "Safe Shutdown from Outside the '
- Control Room." ;
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- The NRC inspectors observed portions of this remote reactor shutdown test 5 l and tietermined that the test was performed in accordance with the approved !
- procedure, Regulatory Guide 1.68 applicable portions of the FSAR, and f I proposed TS. The followinc activities were performed by the NRC i l inspectors
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1 . Attended the pretest briefing held in the control roo l
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j . Verified that communications were established between the auxiliary g shutdown panel, control room, and remote locations in the plant where j manual equipment manipulations occurred, j l . Observed the transfer of equipment and plant control from the control i room to the auxiliary shutdown pane I l . Observed and verified plant cooldown to 350*F and 350 psi ;
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i . Verified the operation of the pressurizer power operated relief (
! valves from the auxiliary shutdown panel by observation of pressure i j drop from the brief opening and closing of the valv {
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Observed transfer to the residual heat removal (RHR) system for !
additional cooldown at a rate that did not exceed TS requirement l j !
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The test equipment was properly installed and calibrated. The procedures !
were available to personnel conducting the test. The test data was !
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collected and recorded in accordance with this procedure. Licensee 1 l i
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personnel were qualified to conduct and record data related to this tes With the exception of the manual charging bypass valve described below, the test procedure was successfully implemente Upon transfer from the automatic charging bypass valve to the manual charging bypass valve, the manual valve failed "shut" after a short period of time. Charging control was returned to the automatic bypass valve and was controlled from the auxiliary shutdown panel. A nonconformance report was issued for the damaged valve and a test change notice (TCN) was issued because manual ch1rging flow control could not be maintained throughout the cooldown as required by the test procedu The remote reactor shutdown system responded to test signals, operated in accordance with established parameters, and remained within calibration requirements. The input signals demonstrated the ability of the remote reactor shutdown system to adequately shut down and naintain the reactor in cold shutdown condition. The electrical independence and redundancy performance functions, operations conducted at the remote shutdown instrument and control panel, including connunications and status and equipment indications, and operation of override control functions were adequately tested and verified to operate in accordance with design requirement The NRC inspector's review of the test procedure and observations durino this test verified that errors previously reported (NRC Inspection Report 50-498/88-01, paragraph 7.3.2, "Conclusions Concerning the Safe Shutdown Drill") in the procedure and training of operators had been correcte The NRC inspectors will review the approved test results when the licensee has completed the evaluation of these test result No violations or deviations were identified in the review of this inspection are . Auxiliary Feedwater (AFW) System Test Unit 2 (70438)
The NRC inspector reviewed Preoperational Test Procedure 2-AF-P-03
"AFW Water Hanmer Test," Revision 0, dated February 2,1988. The purpose of this test was to provide assurance that flow instabilities, such as water hammer, will not occur in AFW system components or piping. The test objective was met under simulated AFW actuation conditions with the SG at the 2/4 low-low steam generat r level, all four reactor coolant pumps running, and the steam dump system in the pressure control mod The NRC inspector observed portions of the valve lineup conducted during preparation for the test prerequisites. The NRC inspector also witnessed the implementation of the water hammer test. The NRC inspector detemined that the test was perfortned in accordance with the approved procedure, the test procedure was available to the personnel conducting the test, the
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l test equipment was properly installed, and the tee. data was collected and recorded in an approved manner. The test procedt..'e was implemented
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correctly and the test results were satisfactor No violations or deviations were identified.
I 11. Control Rod System Test Unit 2 (70432)
The NRC inspector reviewed Preoperational Test Procedure 2-RS-P-02, j "Control Rod Drive Mechanism Operation," Revision 0, dateu March 28, 198 The purpose of this test was to verify and record the current icrofiles provided by the slave cyclers for each control rod drive mechanism (CRDM)
when operated in both the insert and withdraw modes at the maximum operating speed. This test showed that each slave cycler proviced its associated power cabinet with the appropriate cormiand signals to obtain
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the proper CROM timing and signatures during control rod insertion and withdraw operations at maximum spee The NRC inspector verified that the testing was conducted in accordance
- with the approved procedure. During the test, the NRC inspector verified i the cabinet test points for the visir. order recording of Rod "B.B" and the i control bank position and the cabinet group selecc position at Power J Cabinet 1AC. This power cabinet supports three groups of rods with each group containing four rods. The NRC inspector also witnessed the taking of lift, rr. .eable gripper, and stationary gripper current tracings and verified t ,e adequacy of the recorded times for each sequence. The test was implemented correctly and the test results were ' satisfactory.
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No violations or deviations were identifie . Fuel Handling Building Unit 2 (50073)
A final construction inspection of the FHB was performed to assess its
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adequacy for handling new fuel and to verify completeness of construction.
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The inspection consisted of a complete walkdown of every room and system.
1 The following items were noted by the NRC inspector:
j . Bay "A" at elevation -29 feet along the east wall does not fully
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drain because the wall penetration is not at floor leve . High Head and Low Head 51 and Containment Spray suction piping in "B" and "C" train were leaking at the flanges. Flange F0-1423 in Room 205 was elso leakin .
Hand wheel was missing on a 2-inch valve, No. SI-0164
. Valves 51-0101C and DW-0607 had minor seat leak _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _____ _-__-______ __ __ _ - _ - _ _ _ - _ _ _ _ _ _ _ _ _
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. In the primary sample room, Cabinet 9Z5442LP666 had a broken plastic line (Connection 2Y to Hel). Also, in Cabinet 9Z5427LP738, the liquid gross Actuator Meter R15H2519 and Integrator AIT2463 were not installe . Excessive vibration was noted on Safety Valves PSV4612 (Line CC-1441-WA3) and P5V4C10 (Line CC-1440-WA3).
Filter casing leaks (liquid waste) and coupling guard was not
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installed on Pump 7R302NPA-215A in Room WL-001.
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. Heating, ventilating and air conditioning (HVAC) drip pans in Room 002 were stopped up and water was running over into the housing.
! Various miscellaneous items such as left over construction items, missing l cover bolts, excessive lubricant and on going housekeeping items were noted and corrective action was taken during the inspection. It appeared that the FHB was complete and ready for receipt of new fue No violations or deviations were note . Quality Assurance (QA) Program (Test and Experiments) Unit 2 (35749)
The licensee's program for review and evaluation of changes, tests, and d experiments in accordance with 10 CFR 50.59 is described in
- Procedure IP-03.20Q, "Interdepartmental Procedures," Revision 1 dated November 20, 1987. This procedure addressed applicability of 10 CFR 50.59 to control, process, and implement procedures and facility modifications.
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tests, and experiments. The procedure applied to the following:
Safety-related and nonsafety-related modifications
Permanent, temporary, and emergency facility changes
Permanent, temporary, and emergency changes to operations-related and engineering-related procedures
Technical Specifications changes
Software changes ( e., safety-related)
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- The procedure defined appropriate terms, including changes, tests, i
experiments, and unreviewed safety questions. The proceJure defined the responsibilities for reviewing proposed cnanges, tests, and experiments; perfoming required 10 CFR 50.59 evaluations; and developing implementing procedures This procedure described the appropriate actions for the originating group; Plant Engineering Manager; Plant Operating Review Comittee (PORC) as required by Procedure OPGP03-ZA-0004, "Plant Operations Review Comittee," Revision 7; Plant Manager; Nuclear Assurance; and Nuclear Licensing. The procedure included the requirements for retention of records of facility changes, safety evaluations, tests,
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experiments, and arocedure Attachments to the pr;cedure included a flow chart to aid in t1e processing of 10 CFR 50.59 evaluations, guidelines for conplying with 10 CFR 50.59, and a cop / of the form to be used as a record of the evaluation.
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Procedura OPGP03-ZA-0003, "License Compliance Review," Revision 8 dated i
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March 18,1988, estsblished administrative controls for performing License '
Compliance Reviews on plant procedures, instructions, modifications, special tests, and experiments, including changes and other items addressed by STP plant policies, programs and procedure l The 3rocedure required reviews of proposed changes, tests, and experiments
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to tie facility or procedures to determine whether or not a 10 CFR 50.59 <
evaluation is required. The proposed change, test, or experiment nust be
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evaluated to determine whether or not a change to the FSAR, TS, or an
- unreviewed safety question could be involved. A record of the review is ;
required to be maintained. The evaluations must be conducted by personnel ,
with appropriate technical expertise and approved by the department manager. The procedure re.c.' ired that Nuclear Licensing submit an annual "
report to the NRC with a description of changes, tests, and experittents,
- including a sumary of the 10 CFR 50.59 evaluation The licensee's QA program related to the control of changes, tests, and
experiments was in conformance with Regulatory Requirements (Section 6 of i the proposed TS and 10 CFR 50, Appendix 8 Criteria I and X1), licensee
- comitments in FSAR Sections 13.5 and 17.1, and aopropriate industry j guides and standards (ANSI N45.2.8-1975 and ANSI N18.7-1976). [
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) No violations or deviations were identified in the review of this
inspection area.
l 14. Design Changes and Modifications Unit _2 (35744)
The NRC inspector detemined that the licensee had developed and [
1 implemented a QA program related to the control of design changes and I modifications. The licensee's Quality Assurance Plan (QAP) was in [
f conformance with Regulatory Requirements (Section 6 of the proposed TS and l 10 CFR 50.59); licensee commitments (FSAR Section 17.2.3, "Design 1 Control"); ASME B&PV Code,Section XI Articles IWP 3000 and !WV 3000; and
. industry guides and standards (ANSI N45.2.11-1974 and ANSI N18.7-1976). ,
The NRC inspector reviewed STP's design changn and modifications program as described below:
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. FSAR, Section 17.2.3, "Design Control" ,
. Operations Quality Assurance Plan (0QAP) Section 5.0, "Maintenance, h
Installation of Modifications and Related Activities," Revision 1, t
) dated December 2, 1986, and Section 6.0, "Design and Modific1 tion ;
! Control " Revision 2, dated January 30, 1987 l
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. OEEE03-IP-03.010. "Plant Modifications " Revision 3, dated May li, 1988
. The following Operations Engineering Procedures (0EP):
- OEP-1.01Q, "Engineering Organization and Responsibilities,"
Revision 3, dated March 28, 1988
- OEP-1.02Q, "Engineering Approval Authority," Revision 3-CN-1, dated March 28, 1988 [
- OEP-2.02Q, "Design Verification," Revision 3, dated February 15, 1987
- OEP-3.04Q, "Preparation of Modification Evaluation Package,"
Revision 3, dated May 23, 1980
- OEP-3.05Q, "Preparation of Modification Design Package,"
Revision 3 dated May 23, 1988
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- OEP-3.10Q "Modification Closeout," Revision 3, dated May 23, 1988 i OEP-3.13Q "Design Control Program," Revision 3, dated May 16,
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- OEP-6.02Q, "Mainterance and Control of Drawings," I Revision 3-CN-1, dated May 16, 1988
- AEP-6.03Q, "Design Document Chenge Control," Revision 2-CN-2, !
dated June 7, 1988 f r
. Design Criteria Manual, STP, Bechtel Energy Corporation, Volumes I, '
II, !!!, and IV
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. Procedure OPGP03-ZO-0003, "Terrporary Modifications and Alterations,"
Revision 9, dated June 15, 1988 The NRC inspector reviewed the status of implementation of the plant *
design change and modification program. Modifications have not been performed by Nuclear Plant Operations Department (NP0D) or HL&P Engineering. Modifications required during construction have been processed under engineering guideline Modification requests are initiated by completion of a Modification Traveler (MT). The MT is used to process and track the modificatio Check lists are maintained from the modification's conception through installation and testing phases to assure that issues were addresse *
evaluated, and completed. A codification is processed as folicws:
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--______________ ___ _ _-____ _ _ _ _ _ . __ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
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. Modification Evaluation Package - the package identifies and describes the modificetion
. Modification Design Package - the package includes the detailed engineering design, and license compliance reviews, and 10 CFR 50.59 evaluation, if applicabl . Modification Installation Package - the package contains the installation and prerequisite testino information
. Modification Completion Package - the package contains the acceptance test and operability checks This review of the licensee's procedures determined that the edministrative controls and requirements were adequate to control the plant design and modification program. The licensee's program included pro:edure review and approval, verification that TS (proposed)
requirenents were maintained, and that 10 CFR 50.59 evaluations are oerformed. The technical reviews were adequately controlled, approved procedures were used, and the status of designs and modifications was
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tracked from the conceptional stage through the testing and operational
phases.
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No violations or deviations were identified in the review of this inspection area.
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15. Preoperational Test Witnessing Unit 2 (70312)
For each of the preoperational tests witnessed, the HRC inspectors witnessed and reviewed the following:
. The NRC inspectors determined that the proper procedure and most
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recent revision of that procedure was on file. The inspectors
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verified that the procedure used by the test group (crew) was the appropriate revision of the specified procedure and that the test groups (crews) were familiar with the procedural requirements, including the precautions, limitations, and other applicable requirements (i.e. hold points, equipment and/or instrumentation, data taking, expected parameters, etc.).
. The NRC inspectors sssured that test procedure requirements were met by andthe licensee's personnel test personnel, qualifications of eacnincluding member ofminimum test group the test group rew).(c(crew)
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The NRC inspectors verified that procedural prerequisites and initial conottions were me The NRC inspectors reviewed records, including valve lineups, instrumentation calibrations, and line item signoffs by designated test personnel. The NRC inspectors observed and/or monitored instrumentation, eqcipment operation, and personnel actions while tests were being perforre >
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. The NRC inspectors observed that the appropriate plants systems were
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in operation or available for service prior to start of tests, during conduction of tests, and *,ubsequent to completion of test >
. The NRC inspectors verified that equipment / instrumentation was
- properly calibrated and was within the specified calibration time l period . The NRC inspectors verified that the tests were performed in accordance with the appropriate approved procedures and that criteria
, for interruption, repeat, and continuation of testing were specified i in the procedure l
. The NRC inspectors verified that testing data was recorded, that test
discrepancies, unusual events or conditions, unanticipated problems i or conditions, and significant events were docurented.
- , The NRC inspectors observed test group (crew) methers perfoming l procedural steps, recording data, starting and operating equiptent, ,
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and that communications between tests inembers and remote locations
- were adequate. The inspectors determined on a random basis that
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procedural limitations, precautions, and test steps were adhered to during conduct of the various test ,
. The %C inspectors verified that acceptance criteria were stated in each tcst procedure and that the various test group (crew) members were familiar with the acceptance criteria. The NRC inspectors
! verified that the data was recorded as required by the procedure and that the person in charge of each test assured that the data was ,
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recorded, assembled, and transferred to the appropriate group for
! review and evaluatio ,
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The NRC inspectors detemined that temporary modifications -
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(i.e. Jumpers, special equipment, or instrumentation) were installed, '
tracked, and identified by administrative procedure : .
The NRC inspectors indeoendently reviewed and evaluated the test j results and data. The NRC inspectors were cognizant of test i
activities, test results, and plant parameters that could affect (
l specific tests or test condition i The NRC inspectors determined that the test group (crew) members met [
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training requirements specified by the licensee. This training 7 included specific training to assure appropriate knowledge level of ;
the procedure and test requirements. Qualification / training records l
for personnel involved in preoperational testing were available for l
l examination by the NRC inspector I
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l Portions of the following preoperational tests were observed, reviewed and/or evaluated by the hRC inspectors: j j
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Licensee Procedure Procedure Title NRC Inspection Report 2-SH-MS-01 Specific Prerequisite 498;499/88-10 i Test Procedure for SG Secondary Hydro Test l
2-RC-P-01 RCS Cold Hydro Test 498;499/88-16 [
SG-E-01 Breaker Testing 498;499/88-28 2-St-P-02 51 Accumulators 498;499/88-28 2-SI-P-01 SI System Train A 498;499/88-28 2-DG-P-02 Tecting of Standby 498;499/88-35 Diesel Generator t
- 2, Train B v 2-DG-P-02 Emergency-Standby 498;499/88-35 ,
Power Supply System Test 2-CH-P-01 Essential Chilled 498;499/88-41 Water System 2-DG-P-03 DG No. 23 498;499/88-41 2-CV-P-01 CVCS Charging, Let- 498;499/88-41 down & Seal Injection 2-SI-P-02 SI Accumulators 498;499/88-50 2-SF-P-03 Safeguard Test Cabinet 498;499/88-50 Train A 2-SF-P-04 Safeguard Test Cabinet 498;499/88-50 Train B 2-SI-P 04 $1 System Train B 498;499/88-50
?-SF-P-05 Safeguard Test Cabinet 498;499/88-50 Train C 2-SP-P-01 Solid State Protection 498;499/88-50 System (SSPS)
2-SP-P-02 Reactor Protection 498;499/88-50 Master Relay Test 2-HM-P-01 MAB HVAC Systen 498;499/88-50
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2-HE-P-02 Electrical Space HVAC 498;499/88-50 System 2-CH-A-03 MAB Chilled Water 498;499/88-50 System 2-IB-A-01 Loose Parts Monitorirg 498;499/88-55 System Test 2-RC-P-10 Remote Reactor Shut- 498;499/88-55 down Test 2-RS-P-01 Rod Control System 498;499/88-56 2-HM-P-01 Mechanical Auxiliary 498;499/88-56 Building HVAC System 2-HE-P-02 Electrical Penetration 498;499/88-56 Space HVAC System 2-CH-A-03 Mechanical Auxiliary 498;499/88-56 Building Chilled Water System 2-RS-A-01 CRDM Power Supply 498;499/88-56 (MotorDriven Generator Sets)
No violations or deviations were identified in the review of this inspection are . Allegation (Technically Closed) 88-A-0035 (92701)
An allegation was made that an individual in the Physical /Dinensional Laboratory (Met Lab) nay have signed off on calibrations of torque wrenches which he did not perfom or was not qualified to perform also, the alleger was concerned about possible falsification of document An investigation of the allegation by the licensee's SAFETEAM detennined the following: The allegation was directed to nonsafety-related concerns, The alleger, a pipefitter, brought Torque Wrench No. ST-CC-6499 and Adapter No. CNR-0076 to the net lab. The pipefitter indicated that the docurented foot-pcund values for the oial readings and actual torque applied, did not appear normal / correct when compared to other similar reasuring and test equipment (MTE) he had previously use The lab leader concurred that the documentation did not indicate the nomal expected accuracy span for this type instruren ._
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c. The lab leader requested that 6 technician not previously involved in these calibrations (Tech B) recalibrate the wrench and adapter. The recalibration confirmed existing data that indicated the normal expected accuracy span. The recalibration data was completely different from the documented data developed by the technician who originally performed the calibration (Tech A),
d. The torque wrench and adapter had been documented as calibrated by Tech A less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to this recheck. The torque wrench and adapter had been issued to the field, but had not been used, The Fourpipefitter of the sixreturned a total of six were recalibrated by :uspect wrenches a met lab tech (notand adap)ter Tech A on April 29, 1988. During the recalibra'. ions, it was noted that:
. the documentation for four items displayed an improper accuracy span,
. recalibrations showed significantly different data than original documentation, and
. the suspect MTE had all been calibrated and documented less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> earlier by Tech A. The results of re-calibration on the other two items were inconclusive because of differences in calibration of MT f. Met Lab management requested that Tech A demonstrate that he knew the proper procedure and could use proper technique to calibrate torque wrenches and adapters. During this demonstration. Tech A demonstrated that he knew the appropriate procedure and that he could calibrate torque wrenches and adapters properly. The calibration data obtained by Tech A agreed with calibration data generated by Tech 8, who had previously recalibrated this MT g. Tech A could not reproduce his >riginal documented data on these item h. Met Lab management concluded Tech A had knowingly calibrated MTE incorrectly, i. To determine if wrong / incorrect calibrations had been performed previously by Tech A. 5 percent of Tech A's calibration data for the last 6 months was reviewed. Met Lab records indicated that Tech A had performed 894 calibrations during this period. Met Lab management randomly selected 45 items from Tech A's previous calibrations for recalibration verification by other lab tech The recheck on Tech A's previous calibrations indicated that Tech A's prior calibration activities were acceptable. Met Lab managvent concluded that there was no data to indicate that Tech A's prior 1rk wds wrong / incorrect.
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j, Training and qualification records indicated that Tech A met or exceeded the minimum experience required by ANSI N18.1. 4. He was certified to ANSI N45.2.6 as a Level II and was qualified to perfonn calibrations identified on the Level 11 meteorology task qualification training matri Records indicate that while employed at HL&P, Tech A performed 1,149 torque wrench calibrations of which 286 included an adapte . The investigation substantiated the concern that Tech A had knowingly calibrated the MTE incorrectly. The investigation could not detennine whether or not Tech A had falsified calibration docunentation, although evidence was revealed that indicated Tech A had falsified completing required reading assignments, Tech A resigned on May 3, 198 A review of the SAFETEAM investigation substantiated the allegation insofar as torque wrenches and adaptars associated with nonsafety-related matters had been improperly calibrated by a particular technicia No violations or deviations were identifie This allegation is technically close . Exit Interview The NRC inspectors rret with the licensee personnel (denoted in paragraph 1) on Septerber 2,1988. The NRC inspectors sumarized 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 inspectors.