IR 05000498/1988050
| ML20153E583 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/01/1988 |
| From: | Bundy H, Holler E, Hunnicutt D, Mckernon T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20153E546 | List: |
| References | |
| 50-498-88-50, 50-499-88-50, NUDOCS 8809060354 | |
| Download: ML20153E583 (19) | |
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APPENDIX U.S. NUCLEAR REGULATORY C0tEISSION
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REGION IV
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NRC Inspection Report:
50-498/88-50 Operating License:
50-499/88-50 Construction Permit:
CPPR-129
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Dockets:
50-498
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50-499 l
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i Licensee: Houston Lighting & Power Company (HL&P)
i P.O. Box 1700 i
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Houston, Texas 77001 l
Facility Name:
South Texas Project (STP), Units 1 and 2 j
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l Inspection Conducted: July 11-15 and 25-29, 1988
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inspectors:
2X jbmEuN
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D. M. Hunnicutt. Senior Project Engineer Date '
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Project Section 0, Division of Reactor l
Projects i
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0.0ll fft/98
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T. O. McKeffion, Reactor Inspector, Test Da'te /
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l Programs Section Division of Reactor Safety
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c0 H. F. Bundy Reactor Inspector Test Programs DaWe '
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Section, Division of Reacter Safety h
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Approved:
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I E. T. Hbller, Chief Project Section D ~
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Division of Reactor Projects j
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i Inspection Summary
Inspection Conducted July 11-15 and 25 29. 1988 (Report 50-498/88-50)
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Areas Inspected: Routine, unannounced inspection of licensee action on
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previously identified inspection findings, review of licensee event reports.
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review and evaluation of 10 CFR Part 21 Reports, Generic Letter Action Item i
Followup, Three Mile Island Action Item Followup, and observation of planned main generator trip from 100 percent of rated thennal power.
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l Results: Within the six areas inspected, no violations or deviations were
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identified.
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Inspection Conducted July 11-15 and 25-29, 1988 (Report 50-499/88 50)
l Areas Inspected: Routine, unannounced inspection of licensee action on previouslyidentifiedinspectionfindings,reviewof10CFRPart50.55(e)
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l reports (IRCs), Generic Letter Action Item Followup. Three Mile Island Action
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j Item Followup, preoperational procedure review, preoperational test results
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evaluation, and preoperational test program implementation.
Results: Within the seven areas inspected, no violations or deviations were identifie _ - _ _ _ - - -.
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i DETAILS i
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Persons Contacted s
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HL&P t
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J. Westermeier, STP General Manager
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M. Wisenburg, Plant Superintendent Unit 1
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W. Wellborn, Supervising Prcject Engineer
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S. Head Supervising Project Engineer i
K. O'Gara, Project Cornpliance Engineer
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i A. Mikus General Supervisor, Construction
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M. Polishak, Lead Engineer, Project Cornpliance Group
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D. Parker, Startup Engineer
i G. Parkey, Plant Superintendent. Unit 2
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A. Harrison, Supervising Project Engineer
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I J. Bailey, Manager, Engineering and Licensing
j J. Slabinski, Operations Quality Control (QC) Supervisor
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i T. Jordan,ProjectQualityAssurance(QA) Manager
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M. Duke, Staff Engineer
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j J. Geiger, General Manager, Nuclear Assurance
S. Rosen, General Manager, Operations Support
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Bechtel K. McNeal, Project Quality Assurance Engineer f
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R. Moore, Assistant Quality Control (QC) Site Supervisor
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E. Rosoc. Site Manager i
R. Sisson, Senior Resident Engineer P. Phelon, Quality Control (QC) Supervisor j
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D. Hunnicutt, Senior Project Engineer l
J. Bess, Resident Inspector i
O. Garrison, Resident Inspector
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R. Vickrey, Reactor Inspector
l L. E11ershaw, Reactor Inspector W. McNeill, Reactor inspector
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T. McKernon, Reactor Inspector l
H. Bundy, Reactor Inspector l
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P. Wagner, Reactor Inspector l
l All persons listed above attended the exit interview on July 29, 1988.
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4 2.
Licensee Action on Previously Identified Inspection Findings (9t701 and l
92702)
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(Closed) Violation 498/8801-02:
Tempc*ary Hodifications l
The requirements for a temporary modification had not been met in that no markup on a drawing had been made.
The NRC inspector reviewed the cause and corrective actions related to temporary modifications and markup on drawings. The licens,t detertnined that the causes for this violation were an engineer's lack of attention to detail in following the procedure, lack of positive controls on the Temporary Modification Request (THR), and a lack of clarity in the procedure with regard to blank and blind flanges. The licensee updated Drawing SR289F05038 in compliance with Procedure OPGP03-ZO-003, "Temporary Modifications and Alterations,"
Revision 8. dated February 18, 1988. The licensee revised Procedure OPGP03-Z0-003 to include four additional System Engineer responsibilities. The licensee revised Procedure OPGP03-ZE-0031.
"Design Change Implementation After Turnover," Revision 4 to include requirements that Plant Operations and Nuclear Training receive copies of design changes upon approval and that these departments receive notification of design change implementation. This item is closeo.
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(Closed) Violation 498/8801-03:
Locked Valves inadequately locked and/or unlocked valves were not identified and corrected in accordance with the lictnsee's locked valve program.
The NRC inspector reviewed the apparent causes for this violation (failure to adequately lock and/or control valves in accordance with the Locked Valve Program) and the licensee's corrective actions related to Procedure OPGP03-ZO-0027, "Locked Valve Program."
Revision 4 The licensee appropriately identified the locked and/or unlocked valves in accordance with the Locked Valve Program. Valves required to be locked were inspected for proper position and locking devices or implementation of administrative controls.
The valves were observed to be in the required positions end either properly locked or administrative controls were implemented through the South Texas Project Electric Generating Station (STPEGS) clearance process. This item is closed, c.
(Closed) Violation 498/8A01-05:
Surveillanea procedure Discrepancies The licensee failed to follow procedures associated with the Surveillance Program and/or failed to provide adequate procedures to control the activities affecting the quality of surveillance as identified in this violation.
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The NRC inspector reviewed the documentation related to the licensee's investigation and corrective actions. The licensee conducted a special investigation to review the Plant Operating ReviewCommittee(PORC) process;evaluatedprocedures, personnel decisions and actions; revised existing procedures.
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l (Procedure OPGP03-ZA-0002, "Plant Procedures," Rrevision ll, dated
February 29, 1988, and Procedure OPGP03-ZA-0004, "Plant Operations i
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Review Comittee," Revision 9, dated June 20,1988); and trained
personnel in use of procedures.
The licensee's corrective actions l
adequately resolved the potential surveillance procedure discrepancies. This item is closed.
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(Closed) Violation 498/8809-06:
The Licensee Failed to Reestablish I
the Proper Testing Configuration
his violation stated that the licensee failed to reestablish the
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proper testing configuration as required by Station
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Procedure ITEP07-AF-0010 and that this resulted in obtaining im) roper test results. The licensee took exception to the violation witi regard to the performance of Procedure 1TEP07-AF-0010. "Auxiliary
Feedwater Proof Test," Revision 0.
The licensee repeated the test (Procedure ITEP07-AF-0010) to preclude the possibility that the test results would be subject to questinns frcm the NRC. The licensee
agreed that a procedural violation associated with the performance of l
Procedure 1EPO4-ZL-0024 "Rod Drop Time Measurement" had occurred, j
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The NRC inspector reviewed the documentation and corrective actions related to the lack of control associated with the retctor trip
breakers in Procedure 1 PEPO 4-ZL-0024 The lack of control was caused by an inadequete review during the preparation and review of the I
procedure. The licensee's corrective actions included suspension of
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I testing under Procedure 1TEP07-AF-0010 when questioned by an NRC l
l inspector and reperfoming this test, taking into consideration i
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coments regarding the NRC inspector's interpretation of the N9te i
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preceeding Step 4.7 in this procedure. The licensee revised Procedure 1 PEPO 4-ZL-0024 to incorporate specific steps to control the position of the reactor trip breakers. This revision provided f
consistency among various STP Procedures. This item is closed.
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(Closed)OpenItem 498/8812-01:
RCS Flow Measurement Results l
The NRC considered the Reactor Coolant System (RCS) flow measurement results to be an open item pending completion of documented test l
I results and resolution of the actual RCS flow rates with the flow
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rates measured by the short radius elbow differential pressure tans.
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Test Deficiency Record 1 PEPO 4-IA-0001-2, Revision 2, documented the l
investigation of the differences between flow rate measurements using l
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elbow tap transmitters and flow data from other four-loop J
Westinghouse plants and South Texas Project (STP), Unit 1.
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investigation detemined that the measured results were acceptable in
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consideration of the expected accuracy of this type of flow measurement.
FSAR Chapter 14.2.12.3 comitted that the RCS flow would be measured prior to exceeding 75 percent of rated thennal power using calorimetric data.
The licensee agreed to compare this measured flow rate to the mechanical design flow rate.
Subsequently Field Change No. 88-0213, dated February 24, 1988, was issued to i
Procedure IPEP04-ZL-0054, "Reactor Coolant System Flow Measurement at
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Hot Standby " to change the differential pressure (d/p) to flow rate conversion equation derivation in Addendum 1 because of to l
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"non-standard" design of the STP Unit 1 RCS elbows.
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The STP Unit 1 eblow taps were designed to ASME B&PV Code, 1974 Edition - Winter 1975, Code Case 1423-2 Section !!!, NB-3690, which required compliance with NB-3640 for pressure design, NB-3650 for j
pioing analysis, and ANSI B16.11-1966 for socket welded half
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couplings. Functionally, the elbow taps are consistent with the recommendations of ASME B&PV Code as outlined in the text, "Fluid
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Meters."
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Standard Wtstinghouse designed RCS elbows have a radius of 51 inches.
STP Units 1 and 2 RCS elbows have a radius of 37 inches.
The flow
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rate coefficient correction between the standard elbow radfus and STP Units 1 and 2 elbow radius was made. Calculations were completed using the relationship between flow rate and elbow tap d/p obtained c
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from ASME B&PV Code "Fluid Meters," Sixth Edition.
The RCS flow rate measurements and calculations using the correct
elbow radius values at 75 percent of full power and the agreement of l
the flow rates with TS requirements and FSAR comitments verified j
that the RCS flow rates are acceptable.
This item is closed.
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(Closed) Open Item 498/8812-02:
RCS Flow Test Requirenents
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measurements at hot standby lacked the substance necessary to assure j
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The licensee completed the RCS flow measurmnents at 75 percent of i
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full power on July 10, 1988. Station Procedure 1 PEPO 4-ZG-0007, J
' Reactor Coolant System Flow Measurement At Power," Revision 3, dated
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May 16, 1988, verified thennal design RCS flow rate using plant l
calorimetric data. The calorimetric data, steam table properties, and resultant flow rate calculations are documented on Data Sheet i
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System Flow Measurement At Power," Revision 3, dated July 10, 1988
The RCS flow rate test measured the RCS flow rate and verified design (
RCS flow rate using plant calorimetric data prior to operating above
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j 75 percent of full thermal power in accordance with cocmitments stated in FSAR, Chapter 14.2.12.3, Test Description 6. Arendrent 58,
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page 14.2.131.
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l Calorimetric data was obtained using Data Sheet Summary, 1 PEPO 4-ZY-0015. "Statepoint Data Collection," Revision 4, dated July 10, 1988, pages 1 through 4 Data included hot and cold leg temperatures in each of the four loops for the measurements at
75 percent of rated thermal power and subsequent calculations.
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I The relationship between flow rate and elbow tap differential pressure (d/p) is documented in Addendum 1 to Procedure
IPEP04-ZL-0054, Reactor Coolant System Flow Measurement at Hot
Standby," Revision 2, dated October 22, 1987.
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The RCS flow rate measurements at 75 percent of rated thermal power
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and the resultant calculations were compared with acceptance criteria
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stated in the TS and FSAR. The flow rates in gallons per j
ninute (gpm) were as follows:
Calculated Flow FSAR r
l Rate in GPM TS 3.2.5 Minimum TabTe T.1-1
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l Loop at 751 Power Acceptable (GPM)
Minimum (GPM)
1 103,134 98,750 94.100
2 100,486 98,750 94,100
101,178 98,750 94,100
100,157 98,750 94,100 i
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TOTAL 404.955 395,000 382,000 I
l The RCS flow rate measurements, calculations, and agreement between
flow rates stated in the TS requirements and FSAR corrnitments i
verified that the RCS flow rates are acceptable. This item is
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closed.
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(Closed) Violation 498/8833-01: Unexpected loss of Reference
Terperature Signal Trouble shootin of the steam dump controller usin Maintenance Work l
Request (MWR) M -55308 resulted in the unexpected oss of the reactor plant Reference Temperature (Tref) control signal. The signal wat
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lost unexpectedly when instrutrent card TY-660A was pulled as required
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by Step 5.2.5 of the procedure. The Tref signal controls the Rod (
Control (RC) system when the RC is in automatic control.
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l The NRC inspector reviewed the causes and the licensee's corrective i
actions following the loss of Tref.
The aler.a in the control room i
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that would occur during implementation of the MWR, NWR MW 55308 did
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l not specify the system responses that would occur. The loss of Tref l
J signal had no impact on operation of Unit 1.
Card TY-660A was
replaced. WR HS-55308 was completed and Unit 1 plant equipment was l
restored to pre-test conditions.
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b The licensee issued a menorandum on May 14, 1988, that stressed the
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responsibilities of personnel regarding the _need for detailed l
planning and understanding of the actions to be performed and the
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l necessity for adequate consnunication of information to Operations
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sersonnel. The affected Maintenance and Plant Engineering personnel
I 1 ave been trained (course attendance records verifirA training of l
158 personnel).
General Plant Procedure OPGP03-ZM-0021. "Control of Configuration
Changes During Maintenance or Troubleshooting," Revision 0, dated l
June 15, 1988, was preparea, approved, and implemented. This i
procedure provided instructions for the control of configuration changes performed on permanent plant equipment during the f
implementation of MWRs preventive maintenance forms or construction
work request acO vities. This item is closed.
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(Closed) Violation 499/8816-01:
Failure to Follow procedures This violation involved a failure to follow procedures for installing temporary modifications.
During the followup inspection, the RC l
inspector reviewed the licensee s response to the violation, i
ST-HL-?E-2587 dated April 25, 1988, for corrective actions taken.
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addition, the NRC inspector conducted a walkdown inspection of
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applicable relay rooms and logic cabinets. The walkdown inspection verified the licensee's program for control of electrical jumpers and
other temporary modifications is effective and adequate. This item l
is considered closed.
No violations or deviations were identified.
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Review of Licensee Event Reports (LERs)
(90712)
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LER 88-02: Failure to Perform Post Maintenance _ Local Leakage Rate Testing on Containment Isolation valves i
On January 5,1988, with Unit 1 in Mode 5 prior to initial criticality a TS required post maintenance test (PMT) had not been performed on two Containment Isolation Valves (CIV) before entering Mode 4.
Unit I had been operated in Mode 4 after the maintenance work and prior to discovery of the inadequate PMT. The Licensee tested these two CIVs (1-inch ball valves located on the reactor containment building radiation monitor sample exhaust line) and found t'nat one C1Y exceeded its local leakage rate requirements. The CIV was reworked and retesting verified that the local leakage rate was within the specified limits. The licensee revised Procedure OPGP03-ZM-0003, "Maintenance Work Request Program, " Revision 16 dated February 26, 1988. This procedure established a program for reporting and correcting material deficiencies, satisfying the guidelines stated in Regulatory Guide 1.33 "QA Program Requirements (Operations)", Revision 2; FSAR Chapters 13.5, "Maintenance Control" and 3.2, "Classification of Structures, Systems, and Components;"
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Generic Letter 83-028, Item 2.1, (required actions based on Generic
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Implications of Salem ATWS Events); and SER 84-056 (Hispositioning of valves and controls disabled safety systems) and provided
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instructions for processing a Maintenance Work Request (HWR). The
licensee conducted training on the MWR program for shift supervisors
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and support personnel. This LER is closed.
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LER 88-13:
Failure to Perform a Test of the Reactor Coolant System Low Flow Timers as Required by T5 l
On Febnaary 4,1988. Unit 1 was in node 3 prior to initial criticality.
The Itcensee identified two time delay relays in the Solid State Protection System (SSPS) which had not been tested under
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I the surveillance program as required by the TS. This test was
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omitted due to a procedural deficiency.
Licensee actions to prevent recurrence included a review of other surveillance procedures for similar omission, revision to plant procedures on surveillance test i
procedure preparation and review, and revision of surveillance procedures to include low flow timer testing. This LER is closed.
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LER 88-20:
Essential Coolinc Water Screen Wash Booster Pump (ECWSWBP) Histakenly Declarec; Operable
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On February 15, 1988 Unit I was in Mode 5 prior to initial criticality. A review of the inservice test on ECWSWBP 1A revealed i
that the pump test data was outside the acceptable limits.
The pump
was not declared inoperable on February 11, 1988, when the test was
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performed. A new Reference Values Measurement Test (RVMT) was perfonned on February 15, 1988.
This RYMT verified that the pump was
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within the acceptance limits and was operable. Correesive actions included training of licensed personnel using this LER as an example and requiring a second independent review of TS surveillance test results prior to submittal to the Shift Supervisor.
This LER is
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closed, d.
LER 88-25:
Control Roon Ventilation Recirculation Actuation Due To a Radiation Monitor Actuation I
On March 23, 1988 with Unit 1 in Mode 2, an Engineered Safety feature (ESF) actuation of the control room ventilation system to the recirculation node occurred. The Itcensee perfomed diagnos'ic tests on the monitor and attempts were made to duplicate the event. The most probable cause was an inadvertent actuation of a control room ventilation radiation monitor during enaintenance activities. No specific cause was detemined; however, the Itcensee will continue to perform surveillance testing. The licensee completed training for maintenance personnel responsible for testing of radiation monitoring l
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equipment to provide additional assure that maintenance activities are not a cause of future inadvertent actuations.
This LER is
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closed.
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No violations or deviations were identified.
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4.
10 CFR part 50.55(e) Repcrts (IRCs)
(92700)
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(Closed)IkCNo.341(10CFRPart50.55(e)):
Flooding in P9rtions of
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Unit 2 Durinq Heavy Rainfall
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The NRC inspector reviewed the licensee's corrective actions related
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to the flooding in the STP Unit 2 Isolation Valve Cubicle (IVC),
Hechanical Electrical Auxiliary Building (NEAB), and Fuel Handling
Building (FHB). This flooding occurred in Unit 2 only as a result of heavy rainfall (epproxinately 9.5 inches of rain) at STP on i
October 22 and 23, 1986. Mechanical equipment, electrical equipment,
instrumentation, stainless steel piping, and motor operated valves
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j were affected by the flooding.
I Fifty-one nonconformance reports (NCRs) were dispositioned during the
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J cleaning and treggering (if required) of equipment in accordance with
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the various manufacturers' recorrstendations or requirements.
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Equiptrent was evaluated t,y the licensee's engineering staff and the manufacturer (if required) and replaced, refurbished, or repaired, if i
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requi red.
Equiptrent qualification records and documents were
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J validated or recer ted by the manufacturers' for the affected i
components. This n is closed, j
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(Closed)IRC-394(10CFRPart50.55(e)):
Essential Cooling i
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Water (ECW) pump Damage l'
i The NRC inspector reviewed the licensee's corrective actions related l
to damage to the Unit 1 B-train (1B) ECW pump.
The IB pump bearing j
lubrication flow was greater than the flow rate for either the 1A or i
IC pump.
The licensee declared the IB ECW pvep inoperable and
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renoved the purp from service.
The IB ECW pump was subsequently
disasserrbled. The upper and two intertnediate bearings and associated
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l sleeves were extensively damaged. The lower and upper pmp half
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shafts were found to de bent. A 2-inch plastic pipe cap was found
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lodged against the upstream side of the lubrication line orificar
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platt.
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l The IB ECV pump was returned to the purrp manufacturer for refurbishment, including replacement of bearings, shaft sleeves and
j shafts. The manufacturer issued a certificate of corpliance that
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I certified the 1B ECW pump, Serial Nutrber 804402, was repaired in
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i accordance with the requirerents of Purchase Order No. 14926-BF-38055.
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The licensee completed performance and endurance testing for the 1B
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i ECW pump in accordance with NRC IE Bulletin No. 83-05, ASME Nuclear Code Pumps and Spare Parts Manufactured by the Hayward Tyler Pump Company" and in accordance vith Prerequisite Test Change Notice.
"Specific Prerequisite Test Procedure for ECW System Pumps," dated June 12, 1987.
This item is closed.
No violations or deviations were identified.
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10 CFR Part 21 Report Inspections (36100)
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(Closed)10CFRPart21 Report (P21-87-074): Direct Current Motors Wot Qualified by Vendors Program, Direct current motors t. sed to operate motor operated valves (MOVs)
were supplied to some valve manufacturers from Limitorque Corporation
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part of Limitorque Corporation's Qualification Report B-0009, dated I
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April 30, 1976. These motors were manufacturered between December j
1984 and December 1985 by H. K. Porter (now Peerless-Winsmith).
These motors were not part of the Limitorque Corporation's
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Qualification Report because H. K. Porter, with Limitorque
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Corporation's concurrence, had changed design without a formal i
analysis, including the potential effect on environmental
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qualification.
Subsequently, the NRC learned of two significant I
i failures directly attributable to the use of Homex Kapton leads on i
these motors, l
As a result of the Part 21 report and NRC Information Notice B7-08,
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investigation, one of the motors (DIAFMOV-0143)g the licensee's the licensee initiated an investigation. Durin j
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on Unit 1 failed to
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l megger properly and was replaced with a similar notor from Unit 2.
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This deficiency was documented on NCR SE-05724 STP found a total of
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six MOVs with Nomex-kapton leads.
Three of these MOVs had been
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installed in each unit to perform similar applications.
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The licensee reviewed all Unit 1 and Unit 2 safety-related valves
with Limitorque Corporation's motor actuators for the specific
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j nameplate serial number data codes identified in Limitorque
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Corporation's letter of December 19, 1986, to the NRC.
The licensee detemined that all three DC motors in each Unit were associated with i
the turbine driven Auxiliary feedwater (AFW) pump (Train D) of the
AFW System.
These notors are identified as follows:
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Unit 1 Unit 2
i 01AFMOV-0019 D2AFMOV 0019 l
DIAFNOV-0143 D2AFM0V-0143 l
DIAF-FY-7526 02AF-FV 7526
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i Corrective action has been complated.
The licensee has completed a Deficiency Evaluation Report (DER)87-033 to evaluate the reportability of these potential deficiencies.
This DER indicated r
that a safety hazard would not exist and that this deficiency would not be reportable under either 10 CFR Part 21 or 10 CFR f
Part50.55(e).
This item is closed, f
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(Closed)10CFRPart21 Report (P21-87-080):
Containment Hydrogen i
Analyzer Systems This Part 21 Report stated that a design deficiency in the containment hydrogen analyzer systems could permit a loss of alibration gas before the scheduled replacement interval of the storage bottles is recched. Loss of the calibt ation gas would render the systetn inoperable. The hydrogen analyzers were originally i
provided by Exo-Sensor, Inc., which has since been purchased by l
Whitaker, Inc.
An investigation by the licensee verified that the hydrogen analyzers at STP are Model K-!!! manufacturered by Comsip Delphi, Inc. At STP each train's gas supply downstream of the storage bottle consists of i
a pressure regulator, a bubble tight solenoid shutoff valve, and a
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flow control /f solation valve.
Pressure of the gas upstream of the
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solenoid valve is 25 psig. The shutoff valve is rated to 500 psig.
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The licensee's procedure includes a check every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure equipment operability. This item is closed.
c.
(Closed)10CFRPart21 Report (P21-87-081):
Eastern Testing and Inspection Records pertaining to NDE Services This Part 21 report indicated that quality assurance controls including themoceter serial numbers and surface temperatures of the items examined, calibration records, and technicians' eye examination records pertaining to nondestructive examination (h0E) records utilized by Eastern Testing and Inspection, Inc., (ETI) during perfomance of NDE services at Peach Bottom Units 2 and 3 were not l
perfomed as required.
The licensee's investigation detemined that Westinghouse had never i
directly used the services of ET!; however, Westinghouse detemined
that one supplier, Joseph Oat Corporation, had utilized ETI for l
Level I work, specifically shooting radiographs under Joseph Oats
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Corporation supervision and its Quality Assurance Program. Actual
interpretation of the radiographs was conducted by Level !! or
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l Level !!! Joseph Oat Corporation employees. The density and
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correctness of the radiographs were verified by Jospeh Oats Corporation personnel and found acceptchle. Westinghouse concluded that the use of ETI had not affected the safety and/or quality of the components identified by its investigation. The investigation i
detemined that ETI was not on Bechtel's Evaluated Supplier List.
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Bechtel contacted 16 vendors and detemined that only Joseph Oats l
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Corporation had used ET1 for radiographic services on STP orders.
The results of the licensee's investigation demonstrated that ETI
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had not perfomed quality or safety-related work related to the
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reported Part 21 ceficiencies. This item is closed.
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d.
(Closed)10CFRPart21 Report (P2187-083):
Basler Electric Transformers
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I A saturable core transfomer (part No. BE12173-001) manufacturered by Basler Electric in Highland, Illinois, failed in service. An f
inspection by the licensee (TVA) of the failed transfonner revealed t
I that the insulation between the windings was inadequate.
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The licesee performed an investigation and detemined that Basler f
i Electric 7ransforl..er No. BE12173 01 was not used in any of the
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j standby emergency diesel generators at STP. This item is closed.
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e.
(Closed)10CFRPart21 Report (P21-87-084):
Borg Warner Gate Valves
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A potentiel substantial safety hazard related to fasteners installed f
in motor operated 16" x 12" x 16" gate valves manufacturered by Borg
l Warner and supplied by Cochustion Engineering and installed in the i
Shutdown Cooling System (SCS) at Palo Verde Nuclear Generating l
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Station Unit 3 was reported as a Part 21 Report.
The licensee's investigation detemined that Bechtel had purchased
only 2-inch and smaller gate, globe, and check valves from Borg
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Warner. Westinghouse had 60 purchase orders with Borg Warner and f
i suppled no Borg Warner valves to STP.
Therefore, there are no Borg l
Warnsr valves at STP that should be evaluated for potential problems
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reported in this Part 21 report. Thb ' tem is closed.
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l No violations or deviations were iden'
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Generic letter Action Item Followup We
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(Closed)OpenItem 498/8739-04: Genei. etter (GL) 83-23. Item 2.2,
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"Equiprent Classification and Vendor Interrace"
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I The NRC inspector reviewed the licensee's response to NRR letter dated i
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May 4, 1987 The licensee response discussed the implecentation of the f
i Nuclear Utility Task Action Corvnittee/ Vendor Equipment Technical t
Infomation Program (NUTEC/VETIP) at STP and the quality assurance
'l controls over vendor-supplied service on safety related equipnent, i
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The licensee prepared seven pt scedures to provide HL&P (STP) with a l
method of cocriunicating with hRC, INP0, other utilities, and vendors (
j regarding equipeent technical inforration.
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The licensee ccepleted Revision 5 to IP-1.8Q, "Control of \\endor
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Documents," on June 30, 1988.
This program included a periodic
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l contact (interface)withvendorsofsafety-relatedcomponents. The licensee identified and classified the vendor manuals for the key components referenced in the NRR letter dated May 4. 1987.
l Completion of Revision 5 to IP-1.8Q on June 30. 1988, met the
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comitment the licensee made to the NRC and closed reference to
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revision of this procedure in paragraph 4.b of NRC Inspection j
Report 50-493/C8-10.
c.
The licensee's QA program required vendors performing services on safety-related equipment to be listed on the Approved Vendors List (AYL).
Vendor's performing maintenance services under an MWR
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are under the direct responsibility of the HL&P Maintenance i
Department. The MWR requires that maintenance activities on quality-related equipment or systems be performed in accordance with existing procedures and requirements. instrtetion and procedure controis, and related quality requirements, inciduing specifications, as necessaty.
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These approved procedures establish a single program for the receipt, review, status detennination, and distribution of vendor supplied design ar ! technical docuttats and verify periodically that quality-related vendor manuals are current and can serve as a reliable basis upon which the licensee's operation and maintenance may be based.
This item is considered closed for Units 1 and 2.
No violations or deviations were identified, i
7.
Three Mile Island Action item Followup (25565)
(Clesed) Open Item (498/8739-03): TH! Item II.E.4.2. "Containeent
Isolation 3ependability"
The licensee transmitted FSAR changes (HL&P Letter to NRC ST-HL-AE-2182
dated May 29. 1987) describing containment isolation on a Phase B t
isolation signal of the component cooling water (CCW) supply and return to
the reactor coolant pump heat exchangers, reactor coolant drain tank heat exchanger, and the excess letdown heat exchanger.
The CCW flow to the components share comon containment inlet and octiet penetrations.
Additional information concerr.ing containment isolation. CCW supply / return
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to reactor coolant pumps, reactor coolant drain tank heat exchanger and excess letdown heat exchanger was submitted to the NRC by HL&P Letter ST.HL-AE-2237. dated June 11, 1987. HL&P determined that this shared does not meet the requirements of NUREG-0737. Item II.E.4.2(penetrations system arrangement with comon containment inlet and outlet 3)and Standard Review Plan (SRP) 6.2.4. Section !!! which require nonessential systems to be isolated at the containrent on a Phase A containnent isolation signal (safety injection). HL&P requested a deviation from NUREG-0737, Item!!.E.4.2(3)andSRP6.2.4.Section111. The NRC staff considered this deviation from the requirements of NUkEG-0737 Item !!.4.2 and judged that the deviation is acceptable on the basis that l
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adequate isolation capability exists in the form of redundant valves and the piping Jystem itself.
The NRC position is documented in Safety EvaluationReport(SER),NUREG-0781,SupplementNo.4.STPUnits1and2 Section 6. "Engineered Safety Features " paragraph 6.2.4, "Containment
Isolatinn System," July 1987.
The NRC staff position closed this ihm.
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(Closed) Open Item (498/8708-19): _TH! Item I.G.1.3. "Trair.ing
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j Requirements During Low Power Testing"
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The NRC inspector reviewed the following licensee training records to verify that training was provided prior to "hands on" expereince/ training dur W low power testing at Unit 1:
a.
w adure 1. "Control Room Evacuating," POPO4-ZO-0001 Revision 3,
%ed February 1,1988. This procedure evaluated each operating
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crew's performance (licensed and nonlicensed) in the execution of j
the control room evacuation and each individual's performance on the j
assigned watchstations. Drills for each shift were completed during
February 1988.
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Summary of Training for o cification Cycle 6 (January 1 through
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Feb. vary 5,1988). The.-
. of this requalification cycle was h I
prepare operators for the scheduled annual simulator examinations
(during the period between February 15 and March 18.1988) and r.ake available 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per day for self-study for the scheduled written
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I examinations to be administered during this scheduled requalification cycle.
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Course Sumary for Licensed Operator Requalification 701 (conducted during the period between April 20 and Fay 22,1987).
This
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i requalification training included training in off-nortnal procedures, including control room evacuation, relation to postulated fires I
(i.e., control room and relay room firet), and natural circulation l
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rode of reactor control.
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j An Nkt inspector observed a planned shutdown of Unit 1 from a thermal power level of 75 percent and subsequent control of the shutdown reactor
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J from the Unit 1 Auxiliary Shutd0wn Panel (ASP). The shutdown reactor was
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in natural circulation mooe while being controlled from the Unit 1 ASP.
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This scheduled Unit I test verified that operators had received the j
required training. The operators demonstrated their proficiency by perfortning the required functions at the ASP. The operators denonstrated j
that they understood the prt,cedures and that they could implement these
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i procedures, establish and maintain control of Unit 1, and maintain the plant in a safe shutdown condition from outi;ide the Control room using the i
i equipeent and instmentation located at the ASP.
This test was perforced j
in accoNance with specified procedures and licensee comitments, i
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The licensee has comitted to complete training of Unit 2 operators (licensed and nonlicensed) on the ASP by November 23, 1988.
This item is closed for Unit 1 and Unit 2.
No violations or deviations were identified.
8.
Generator Trip From 100 Percent of Full Power (72580)
The NRC inspectors observed the scheduled Unit 1 main generator trip from a steady state power level of 100 percent of rated thermal power. The purpose of this plant trip from 100 percent was to verify the ability of Unit 1 plant to sustain a trip of the main generator and to determine the overall response time of the reactor coolant hot leg resistance temperature detectors (RTDs).
The NRC inspectors observed the tripping (opening) of the main generator output breaker from the Unit 1 control room at 10:33 a.m. (CDT) on July 28, 1988. The NRC insper+. ors observed the plant responses and licenseepersonnelactionstothisnetlossofelectricalload(lossof the electrical load results in the maximum credible overspeed condition for the main turbine) from vantage positions in the control room and on the main turbine / generator deck.
The NRC in pectors determined that this test was cor. ducted in accordance with approved Procedure 1 PEP 04-ZY-0102, "Plant Trip for 100% Power,"
Revision 3, dated July 27, 1988.
During the test period pertinent plant parameters were recorded by licensee personnel.
The recorded data and observed equipment responses verified that associated plant equipment, instrumentation, and components performed in accordance with design requirements and within the anticipated limits. Observations and review of data indicated no major problems or potential nuclear safety concerns.
Observations verified that no excessive vibrations in piping or components occurred during this test. TFe main steam valves closed smoothly and with anticimted force. No unusual noises or equipment malfunctions were observ'd. The minimum reactor coolant system average temperature (Tavg)
e was 558'F Tavg immediately prior to the scheduled plant trip was 567'F.
The NRC inspectors verified that the following acceptance criteria stated in Procedure 1 PEP 04-ZY-0102 was met:
a.
The safety limits stated in TS 2.1.1 and 2.1.2 were not exceeded.
b.
The neutron flux dropped to less than 15 percent of full power value in less than two seconds.
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All control and shutdown rod cluster control assemblies (RCCAs)
dropped into the reactor core.
d.
Safety injection actuation did not occur.
e.
Pressurizer safety valves did not lift.
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Main Steam safety valves did not lift.
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The overall hot leg RTO response time was less than 7.531 seconds.
The NRC inspectors observed the following safety-related conditions and licensee actions:
a.
Plant electrical loads transferred as designed.
Conditions of TS 3.8.1.1 were met during the test.
b.
The turbine bypass valver operated to maintain the RCS within established limits.
c.
Licensee personnel performed necessary manual functions to maintain safe plant limits, including manually tripping the reactor.
In accordance with Procedure 1 PEP 04-ZY-0102, a manual scram was initiated when an automatic scram signal was not received within 2 seconds after opening the main generator breaker.
d.
Licensee personnel placed and maintained the plant in normal shutdown condition (Hot Standby - Mode 3) following the plant trip.
The NRC inspectors determined through observations, discussions with licensee personnel, and review of procedures, TS, and FSAR Chapter 14.2.12.3.23, Amendment 56, that the licensee met the requirements and commitments related to the plant trip from 100 percent of rated themal power. The successful completion of this test verified the ability of the plant and licensee personnel to sustain a trip of the main
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generator from 100 percent of rated thermal power and maintain the plant in a safe condition.
No violations or deviations were identified.
9.
Preoperational Procedure Review (70354)
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During the inspection, the NRC inspector reviewed the following preoperational test procedure:
2NI-P-01, Revision 0, "Nuclear Instruments System," dated July 6,1988 The objectives of this procedure are to demonstrate operability of the Nuclear Instrumentation system, verify each channel functions properly and provides the as-designed output pemissive and reactor trip system interlocks. Within the scope of this inspection, the NRC inspector verified that the procedure was satisfactory and the stated objectives were delineated in the test procedure.
No violations or deviations were identifie,.
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10.
Preoperational Test Results Evaluation The NRC inspectors verified that the results of the following tests were within the stated acceptance criteria and that any deviations were properly dispositioned.
The licensee effectively identified those areas requiring test exception and completed the necessary test procedures and retesting changes The licensee's compliance with administrative controls for test execution and test results review and evaluation was complete and adequate. Changes made to test procedures were properly reviewed, approved, and the procedures annotated. Tests were properly conducted with the appropriate individuals initialing and dating the procedural steps along with the necessary quality assurance signoffs. The NRC inspectors verified that the licensee was meeting the comitments of Regulatory Guide 1.68 and the FSAR.
The licensee's compliance with administrative practices of Startup Administrative Instructions (SAI) 18 Revision 7, "Preoperational Testing," and SAI-19, Revision 6. "Acceptance Testing" was both evident and adequate.
Furthermore, the NRC inspectors verified that all test data met the stated acceptance criteria.
a.
Engineered Safety Features System (70322)
2-SI-P-02, Revision 1. "Safety Injection Accumulators"
2-SI-P-04, Revision 1. "Safety Injection System Train B"
2-SF-P-03, Revision 2, "Safeguard Test Cabinet Train A"
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2-SF-P-04, Revision 1, "Safeguard Test Cabinet Train B" 2-SF-P-05, Revision 1, "Safeguard Test Cabinet Train C"
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b.
Reactor Protection System (70325)
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2-SP-P-01, Revision 0, "Solid State Protection System (SSPS)-
l Reactor Protection Logic Test" 2-SP-P-02, Revision 0, "SSPS - Reactor Protection Master Relay
Test" 2-HM-P-01, Revisior, 1, "MAB HVAC System"
2-HE-P-02, Revision 0, "Electrical Space HVAC System"
i 2-CH-A-03, Revision 1. "MAB Chilled Water System"
I No violations or deviations were identi':ed,
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11.
Preoperational Test Program Implementation (70302)
i During this portion of the inspection, the NRC inspectors verified that the licensee has implemented and complied with written administrative controls over the preoperational testing program.
The NRC inspectors
conducted interviews with the test program director and other testing
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personnel, reviewed a sampling of tests from the test program index, and reviewed qualification records of key test personnel.
The NRC inspectors verified that the test program director was familiar with the responsibilities of key test personnel, lines of authority and
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responsibility and interfaces amongst those organizations involved in the test program. Test procedures were reviewed and approved in accordance with the applicable administrative procedures.
Furthennore, procedures conained references to the most current issues of drawings and vendor's manuals. Component configuration packages (CCPs) were reviewed, processed, and implemented in accordance with procedural controls.
In those instances that warranted retest, the test procedures were properly re'tised to incorporate test of the design changed system. The NRC inspectors conducted interviews with key test personnel to verify their familiarity with administrative controls covering the conduct of corrective and preventive maintenance during preoperational testing.
Furthermore, the NRC inspectors reviewed trainirg records to verify appropriate certification of key test personnel, training had been conducted covering administrative controls for testing, and other applicable quality assurance / quality control indoctrination. Within the scope of this inspection, the NRC inspectors confirmed the licensee's compliance with Regulatory Guide 1.68, FSAR comitments, and guidances provided in ANSI N18.7-1976 and Regulatory Guide 1.58.
No violations or deviations were identified.
12.
Exit Interview The NRC inspectors met with the licensee personnel (denoted in paragraph 1) on July 29, 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.
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