IR 05000498/1987011
| ML20205C549 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/18/1987 |
| From: | Hildebrand E, Hunnicutt D, Greg Pick, Stewart R, Vickrey R, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20205C445 | List: |
| References | |
| 50-498-87-11, 50-499-87-11, IEB-73-06, IEB-73-6, IEB-74-08, IEB-74-8, IEB-75-03, IEB-75-3, IEB-77-03, IEB-77-3, IEB-78-01, IEB-78-1, IEB-79-14, IEB-82-02, IEB-82-2, IEB-85-002, IEB-85-2, IEB-86-004, IEB-86-4, IEC-76-03, IEC-76-3, IEC-77-06, IEC-77-6, IEC-78-02, IEC-78-2, IEC-79-04, IEC-79-4, IEC-80-02, IEC-80-2, IEC-81-02, IEC-81-2, IEIN-79-19, IEIN-84-18, IEIN-85-067, IEIN-85-67, NUDOCS 8703300210 | |
| Download: ML20205C549 (10) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMISSION
REGION IV
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NRC Inspection Report:
50-498/87-11 Construction Permits:
CPPR-128 50-499/87-11 CPPR-129 Dockets:
50-498 50-499 Licensee: -Houston Lighting & Pcwer Company (HL&P)
P. O. Box 1700 Houston, Texas 77001 facflity Name:
South Texas Project,- Units:1 and 2 Im;retion At:
South Texas Project, Matagorda County, Texas In;p?c2 ion Conducted:
March 2-6, 1987 Inspectors:
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3//B/B7 D. M. Hunnicutt, Chief, Operations Section Date
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Reactor Safety Branch h
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/p R. C. Stewart, Reactor Inspector, Operations Date '
Section, Reactor Safety Branch d
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p E./P.)Hildebra6d, eactor Inspector Date
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O Ophp6tions Secti Reactor Safety Branch V
3/ielw ckrh, React @ Inspector, Operations Date
tion, React V Safety Branch 8703300210 870324 PDR ADOCK 05000498 G
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b $1f } W k 3//8/97 G. A. Pick, Reactor Engineer, Operations D'a te '
Section, Reactor Safety Branch Approved:
g/jp/gy T. F. Westerman, Chief, Reactor Safety Branch 04te '
Inspection Summary Inspection Conducted March 2-6, 1987 (Report 50-498/87-11; 50-499/87-11)
Areas Inspected:
Routine, unannounced inspection of licensee reported significant construction deficiencies (10 CFR 50.55(e)) and responses to NRC Inspection and Enforcement Bulletins, Circulars, and Information Notices.
Results: Within the four areas inspected, no violations or deviations were identified.
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DETAILS
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1.
-Persons Contacted
- *M. R..Wisenburg,. Deputy. Project Manager
.*S. M. Head, Lead Engineer Licensing Other
- C. E.' Johnson, Senior' Resident Inspector The NRC inspectors 1also= interviewed other licensee employees during the course of the.. inspection.
- Denotes those~ attending the exit interview.
2.
Review of IE Bulletins The NRC inspectors reviewed the below listed Bulletins issued by the NRC'
0ffice of Inspection and Enforcement.
The actions taken by the licensee were determined to be commensurate with Lthe requirements of the Bulletin and each listed item is therefore considered closed.
Document No.
Title
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IEB 73-06 Inadvertent Criticality in a Boiling Water Reactor IEB 74-08 Deficiency in(ITE-Molded Case Circuit o
Breakers HE-3 IEB 74-12 Incorrect Coils in Westinghouse Type Steam Generator Relays at Trojan IEB 74-13 Improper' Wiring on General Electric Motor Control ~.2 Centers.at Fort Calhoun ssi IEB 75-03 Incorrect Lower Disc Spring and Clearance
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Dimension-in Series 8300 and 8302 ASCO Solenoid Valves IEB 75-06 Defective Westinghouse Type OT-2 Control Switches 2IEB 75-07 Effects of Corrosive Emissions on HEPA Filte ;
IEB 77-03-On Line Testing of the Westinghouse Solid State Protective System (WSSPS)
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IEB 77-04 Calculational Error Affecting the Design Performance of a System For Controlling pH of Containment Sump Water Following a LOCA IEB 77-06 Potential Problems with Containment Electrical Penetration Assemblies IEB 77-07 Containment Electrical Penetration Assemblies at Nuclear Power Plants Under Construction IEB 78-01 Flammable Contact-Arm Retainers in GE CR120A Relays IEB 78-02 Terminal Block Qualification IEB 79-14 Seismic Analyses for As-Built Safety-Related Piping Systems IEB 79-15 Long Term Operability of Deep Draft Pumps IEB 79-17 Pipe Cracks in Stagnant Water Systems at PWR Plants IEB 82-02 Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundry of PWR Plants IEB 85-02 Undervoltage Trip Attachments of Westinghouse DB-50 Type Reactor Trip Breakers IEB 86-04 Defective Teletherapy Timer that May Not Terminate Treatment Dose No violations or deviations were identified.
3.
Review of IE Circulars and Information Notices The NRC inspectors reviewed the below listed Circulars and Information Notices issued by the NRC Office of Inspection and Enforcement.
The actions taken by the licensee were determined to be commensurate with the requirements of the Circulars or the Information Notices and each listed item is therefore closed.
Document No.
Title IEC 76-03 Radiation Exposures in Reactor Cavities IEC 76-06 Stress Corrosion Cracks in Stagnant Low Pressure Stainless Piping Containing Boric Acid Solution at PWR's
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IEC 77-06 Effects of Hydraulic Fluid on Electrical Cables IEC 77-10'
Vacuum Conditions Resulting in Damage to Liquid Process Tanks IEC 77-11 Leakage of Containment Isolation Valves with Resilient Seats IEC 78-02 Proper Lube Oil for Terry Turbine IEC 78-05 Inadvertent Safety Injection During Cooldown IEC 78-07 Damaged Components of a Bergen-Patersen Series 2500 Hydraulic Test Stand IEC 78-15 Tilting Disc Check Valves Fail to Close With Gravity in Vertical Position IEC 79-04 Loose Locking Nut on Limitorque Valve Operators IEC 79-09 Occurrence of Split or Punctured Regulator Diaphrapms in Certain Self-Contained Breathing Apparatus IEC 80-02 Nuclear Power Plant Staff Work Hours IEC 80-03 Protection From Toxic Gas Hazards IEC 80-04 Securing of Threaded Locking Devices on Safety-Related Equipment IEC 80-10 Failure to Maintain Environmental Qualifications of Equipment IEC 80-11 Emergency Diesel Generator Lube Oil Failure IEC 80-12 Value-Shaft-to-Actuator Key May Fall Out of Place When Mounted Below Horizontal Axis IEC 80-14 Radioactive Contamination of Plant Demineralized Water System and Resultant Internal Contamination of Personnel IEC 80-22 Confirmation of Employee Qualifications IEC 81-02 Performance of NRC Licensed Individuals While on Duty IEC 81-04 The Role of Shift Technical Adviser and Importances of LER
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IEC 81-12
.InadequatePeriodic[TestProcedureofPWRProtection
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System IEC 81-13
' Torque Switch Electrical Bypass Circuit for-
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Safeguard Service Valve Motors IEC 81-15'
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' Unnecessary Radiation Exposures to the Public and-Workers During Events Involving Thickness and Level Measuring Devices e
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Pipe Cracks in Stagnant Borated Water Systems
'IEN 79-19 at PWR Plants-IEN 84-18
_ Stress Corrosion Cracking in. Pressurized Water Reactor Systems
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IEN 85-67 iValve-Shaft-to-Actuator' Key May Fall Out of Place When Mounted Below Horizontal Axis
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No violations or deviations were ident'fied.-
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4.
-Followu) on Licensee Reported Significant Construction Deficiencies (10 C:R 50.55(e))
(Closed) Incident Review Committee Item No. 0082 This item involved a finding-that Hilti Kwik-Bolt Expansion Anchors could have reduced load capacities (ultimate' pullout performance). An onsite construction survey identified nine cases of installed Hilti Kwik-Bolts in
pipe support assemblies that were in doubt of meeting specifications.
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test program was conducted by Brown and Root and an independent test agency. The revised ultimate values for Hilti Kwik-Bolt Expansion Anchors.
were reviewed in shear for 1/4", 1/2", and 5/8"~ diameters andJin tension and shear for l diameter by_ the manufacturer. The revised' values were
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verified, as required.
This item is considered closed.
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(Closed) Incident Review Committee Item No. 0087 This item involved a finding that il potential bias could be introduced in-the_ steam generator level instrumentation due to heating of the reference leg following a high energy line break.
A field change notice was issued
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that installed the use of the steam generator narrow range water level l
compensation system which is an integral part of the Qualified Display and Processing System (QDPS).
This item is considered closed.
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(Closed) Incident Review Committee Item No. 0104 This-item involved a finding that failure of some Rockbestos coaxial cables would accrue during aging at elevated temperatures.
Replacement cables were obtained and used that had revised aging and test sequence and n
- design change of~the braid shield.
This item is considered closed.
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(Closed) Incident Review Committee Item No. 0127 This item involved the design of the Auxiliary Feedwater System (AFWS).
The licensee identified the following three areas that required resolution:
a.
Condensate Storage Tank - resolved by installation of a loop seal in lieu of two Safety Class vacuum breakers.
The vacuum breakers were retained and placed in the category of nonsafety vacuum breakers.
b.
Four AFWS recirculation lines outside the Isolation Valve Cubicle (IVC) - Resolved by upgrading these lines to Safety Class 3, Seismic Category 1 piping.
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Pump discharge piping for AFWS designed to lower pressure hydrostatic pressure rating than electric and turbine driven pumps maximum pressure rating.
Portions of the piping have been replaced.
The turbine driven pump and AFWS electric pump have been hydrostatically tested as required by ASME B&PV Code,Section XI.
The piping, AFWS electrical pump and the turbine driven AFWS pump meet the pressure requirements for the design and operation of the AFWS.
This item is considered closed.
(Closed) Incident Review Committee Item No. 0129 This item involved a finding of an undetectable test switch failure in the solid state protection system.
A field change notice was used to make associated wiring changes which assured that test switch failure could be detected.
This item is considered closed.
(Closed Incident Review Committee Item No. 0144 The NRC inspector reviewed the closeout package concerned with this reportable deficiency. The licensee replaced each of the 480-volt engineered safety features impedance with transformers rated at four percent impedance.
This corrective action was initiated to prevent the voltage output of the transformer from dropping below 80 percent of the rated motor capacity whenever large loads such as the residual heat removal (RHR) pumps and/or reactor containment fan cooler (RCFC) were added to the transformer output.
This item is considered closed.
(Closed) Incident Review Committee Item No. 0145 This item involved the finding of a manufacturing discrepancy relative to the undervoltage attachments on DS-416 reactor trip switchgear.
A field change notice was issued that corrected the manufacturing discrepancy to the reactor trip breakers.
This item is considered close "
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(Closed) Incident Review Committee Item No. 0151 This item involved a finding that contact bounce had been experienced in the ' mercury relay of the temperature channel test of process control and protection system.
Replacement relays were developed and installed with a field change notice. This item is considered closed..
(Closed) Incident Review Comittee Item No. 0152 This item involved a finding that some nonqualified instruments were used for class IE applications.- The associated transmitters were replaced with qualified transmitters. This item is considered closed.
(Closed) Incident Review Comittee Item No. 0188 This item involved an isolated error in the analysis of containment
' spray (CSS) pH resulting in a condition whereby spray pH cannot be maintainad less than equipment qualification limits during all modes of CSS operations. A reanalysis has subsequently concluded that the CSS spray effects on equipment qualification and offsite doses are within acceptable limits.. The results of the reanalysis are contained in the update of the FSAR, Amendment No. 56. This item is considered closed.
(Closed) Incident Review Committee Item No. 0189 This item involved time. delays and effects of Safety Injection (SI) Pump starts following a postulated loss of offsite power (LOOP).
In the case of an SI signal imediately followed by a LOOP or a LOOP imediately followed by an SI, the timing of the bus strip and initiation of the loading sequence results in the first and second loads being applied to the diesel generators without a sufficient intervening time delay. A safeguards analysis evaluated the effect after a modification to the Engineered Safeguards Feature (ESF) Load Sequencer. The analysis indicated that the change to the ESF Load Sequencer Logic Design would-have no impact to the consequences of the Loss of Coolant Accident (LOCA)
j analysis. This item is considered closed.
(Closed) Incident Review Comittee Item No. 0227 l
The NRC inspector reviewed the corrective action taken in response to the i
presence in a safety-related system of nonclass 1E level switches. The licensee chose to remove these switches and cause an annunciator to sound
in the control room on low component cooling water (CCW) surge tank level.
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l There currently exists safety grade level transmitters that provide level i.
indication for each CCW train in the control room. This item is considered closed.
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(Closed) Incident-Review Committee Item No. 0229 The reported condition involved the possibility of a small break LOCA due to postulated failure of the reactor coolant pump (RCP) thermal barrier,
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concurrent with single active failure of CCW return isolation valve.
NRC inspector reviewed the closecut binder for this reported deficienc The The licensee installed two self-activated pressure regulated valves in the CCW outlet from each RCP.
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pressure caused by leakage of reactor coolant system water.The valve considered closed.
This item is (Closed) Incident Review Committee Item No. 0232 This item involved a finding of cutting of cable insulation in electrical penetration assemblies.
Inspection and repair of the associated assemblies and design change packages were completed. This item is considered closed.
_(Closed) Incident Review Committee Item No. 0233 in fabrication of HVAC supports.This item involved identification of crac Portions of the material from one heat,The suspect independent laboratory.
number 81191, were identified as unacceptable by the independent laboratory.
from heat number 81191 and all material that could not be from one of the three acceptable heat numbers was rejected and replaced with acceptable material.
preclude recurrence of similar material problems. Recurrence control measures closed.
This item is considered (Closed) Incident Review Committee Item No. 0241 This item involved AFWS pipe spools the,were supplied with incorrect wall thickness.
These 52 pipe spools were 3" and 4" diameter schedule 80 instead of schedule 140.
The AFWS spools were replaced with new pipe spools in accordance with existing procedures and piping specifications.
This item is considered closed.
(Closed) Incident Review Committee item No. 0294 This item involved construction damage to electrical penetration assembly (EPA) conductors.
and/or cut insulation) to EPA cables were identified. Seventeen cases of dama was developed and 6 out of 9,000 safety-related conductors were identified A test procedure as unsatisfactory. Corrective action wac completed.
The vendor manual was revised to require use of temporary slip-tubes to preclude damage to EPAs during installation. This item is considered closed.
(Closed) Incident Review Committee Item No. 0336 on Limitorque Motor Operated Valve (MOV) actuators whic failure of the operators.
Inspection and replacement of tripper fingers
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for the related MOV actuators was accomplished.
Procedure revisions, associated training, and issuance of-a vendor control procedure were also-done. -This item is considered closed.
No violations'or deviations were identified.
5.
Exit Interview Th'e NRC inspectors met with the licensee personnel and the NRC Senior Resident Inspector (Construction) denoted.in paragraph 1 on March 6,1987,'
and summarized the scope and findings of the inspection.
The licensee.
acknowledged the findings.
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