ML20215D795
| ML20215D795 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 06/10/1987 |
| From: | Ellis J Citizens Association for Sound Energy |
| To: | Bloch P, Jordan W, Mccollom K Atomic Safety and Licensing Board Panel |
| References | |
| CON-#287-3778 OL, NUDOCS 8706190147 | |
| Download: ML20215D795 (61) | |
Text
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214/9k6-926 (CITIZENS ASSN. FOR SOUND ENERGY)
'87 JtJN 15 A9 :43 June 10, 1987 0FFU -
np r ' UJ DOCK! ; C Administrative Judge Peter B..Bloch Dr. Kenneth A. McCollom U. S. Nuclear Regulatory Commission 1107 West Knapp Street Atomic Safety & Licensing Board Stillwater, Oklahoma 74075 Washington, D.C.
20555 Dr. Walter H. Jordan 881 W. Outer Drive Oak Ridge, Tennessee 37830
Dear Administrative Judges:
Subj ect:
In the Matter of'
. Texas Utilities Electric Company, et al.
Application for an Operating License Comanche Peak Steam Electric Station Units 1 and 2 Docket Nos. 50-445 and 50-446 - 8 0 Potential 10 CFR 50.55(e) Items To assist the Board in its desire to be kept up-to-date on matters of potential significance, CASE is attaching copies of the potential 50.55(e) items (SDAR's, or Significant Deficiency Analysis Reports) which we have received since we provided the Board with such items on April 30, 1987.
It is our intention to periodically continue to provide these SDAR's unless the Board indicates otherwise.
Respectfully submitted, CASE (Citizens Association for Sound Energy) q i
&cYJ i
, -( rs.) Juanita Ellis President 1
cc:
Service List, with Attachments 8706190147 870610 PDR ADOCK 05000445 9
PDR Q3
m
- r SDAR~INDEX-05-18-87 y
i
'n-(SUB5CT ~
DETERM.
STATUS 1975-REACTOR BUILDING EXCAVATION R
'1975' REACTOR CAVITY MAT R
TXX-1198 l
1976
-UNIT.NO. 1 CONTAINMENT BASE MAT R
1976 SSI' DAM FILTER "A" MATERIAL R
TXX-2069 77-A POLAR CRANE SUPPORT GIRDER' NR NONE.
77-B FUEL' BLDG. DESIGN DISCREPANCY R
TXX-2609' C FAILURE OF WESTINGHOUSE AR/ARLA LATCHING RELAYS R
TXX-2465 77-D POTENTIAL PROBLEM WITH OPERATION OF SAFEGUARDS ACTUATION BLOCK / RESET CIRCUITRY R*
TXX-2481 77-E SPECIFICATION NONCONFORMANCE WITH THE PSAR NR NONE 77-F STATOR WINDING DEFECT <DELEVAL DG' NR NONE.
77 FALSIFICATION OF CONCRETE AIR ENTRAINMENT. RECORDS NR.
.TXX-2266' 77-2 CONCRETE CURING BLANKET FIRE
.NR TXX-2265 77-3 INSTALLATION OF PIPE WHIP RESTRAINT:
-ANCHOR BOLTS, NUTS, & PLATES NR TXX-2474 77 POLAR CRANE SUPPORT BRACKETS R'
TXX-2395 77-5 OMISSION OF BORON RECYCLE EVAPORATOR PACKAGE FOUNDATION ANCHOR BOLTS NR NONE 77-6 OMISSION OF REINFORCING' STEEL FROM CONCRETE POUR IN RX. BLDG. NO. 1 NR NONE-77-7 POLAR BRANE GIRDER UPLIFT RESTRAINTS-NR TXX-2464 77-8 INSTALLATION OF PIPE WHIP REATRAIN
' ANCHOR BOLTS,~ NUTS, & PLATES NR TXX-2474 i
77-9 3 INCH DIAMETER HOLES DRILLED IN R.B. NO. 1 NR NONE 77 OMISSION OF. REINFORCING STEEL FROM R.B.1 & A.B.1 NR NONE 77-11 OMISSION OF REINFORCING STEEL FORM R.B.1 & SG.B.1 NR NONE 77-12 ORIENTATION OF CADWELD SLEEVES NR TXX-2678 1977 CADWELD TEST. SPECIMEN REBAR FAILURE NR TXX-2269
<I 78-A' DAMAGE TO DELAVAL DIESEL GENERATOR CONTROL PANEL NR NONE 78-B MISCLASSIFICATION OF SAFETY RELATED EQUIPMENT NR TXX-2867 78-C SERVICE WATER INTAKE PUMP SEISMIC RESTRAINT NR TXX-2905 78-1 SERVICE WATER PUMP SUPPORTS NR NONE
' 2.
CATALYTIC HYDROGEN RECOMBINERS NR NONE 78 CANCELLED-78-4 ANCHOR BOLTS-MISCLASSIFICATION NR SEE 78-B 78-5.
ANCHOR PLATE WASHERS MISCLASSIFICATION NR SEE 78-B 78-6 DEFECTIVE CADWELDING NR TXX-2896 78-7 WELDING ACTIVITIES FOR UNIT 1 RCS LOOP PIPING NR TXX-2926
'1975-1978:
8 REPORTABLE SDARS I
b
SDAR INDEX-05-18-87
~
r a
'ILO,.
- SUBJECT DETERM. STATUS 79-01 EXPANSION JOINT-DECONTAMINATION ROOM NR TXX-2952
.79-02 INSTALLATION OF PIPE ELBOWS IN SG LOOP PIPING NR TXX-2957
.79-03 REACTOR VESSEL N0ZZLE BASE METAL DEFECTS NR TXX-3007 79 PIPE SUPPORTS ~
NR TXX-3036 79-05 PIPE' WALL THICKNESS R
TXX-3138 79A-PRESSURE-TRANSMITTERS NR TXX-3003 79B POSSIBLE COOLING PROBLEM WITH DIESEL GENERATORS NR TXX-3054
'79-06 ELECTRICAL CABLE TRAY HANGERS-PARTIAL KELDS R
-TXX-3067 79-07 REACTOR CONTAINMENT NR 10/29/79 79-08 INSTALLATION OF DRILLED-IN EXPANSION ANCHORS R*
TXX-3125 79-09 ~ WELDED CONDUIT SUPPORTS R
TXX-3173 79-10 CLASS V PIPING SUPPORTS R
'79-11 REACTOR COOLANT NR ~
12/29/79
.79-12 FILTER INTERNALS NR 01/12/80 79-13 CONCRETE HONEYCOMBING-UNIT 2 SG COMPAR'IMENT R
TXX-3124 4
1979: 15 TOTAL SDARS--6 REPORTABLE 80-01 SERVICE WATER PUMPS NR 01/17/80 80-02 SERVICE WATER CONTROL VALVES R
TXX-3114 80-03 CONTROL BOARD WELDING CONNECTIONS R
TXX-3131 80-04' REACTOR VESSEL N0ZZLE SAFE-ENDS NR TXX-3154 80-05 USE OF ARCHITECTURAL' CONCRETE IN FLOOR SLABS R
TXX-3247 80-06 UPPER INTERNALS ROTO-LOCK INSERT NR~
TXX-3217 80-07 PLASCITE 7122 COATING NR' TXX-3229 80-08
. THREAD ENGAGEMENT OF SWAY STRUTS NR
'TXX-3219 80-09 DIESEL GENERATOR PIPE SUPPORTS R*
TXX-3415 80-10 HILTI-KWIK CONCRETE ANCHOR BOLTS R*
TXX-3442 80-11 CANCELLED 80-12 PULL-0UT PERFORMANCE OF HILTI-KWIK BOLT ANCHORS NR TXX-3398 i
1980: 11 TOTAL SDARS--5 REPORTABLE 81-01 PIPE SUPPORT BASEPLATE GROUTING MATERIAL NR TXX-3279 81-02 ELECTRICAL CABLE LUBRICANT NR TXX-3357 81-A WESTINGHOUSE GATE VALVES R
TXX-3409 81-B BAHNSON HVAC ANCHOR BOLTS R
TXX-3329 81-03
'HVAC COOLING SYSTEM R
TXX-3349 81-04 FAILURE OF 6.9KV CIRCUIT BREAKER NR TXX-3313 81-C VOLUME CONTROL TANK LEVEL CONTROL SYSTEM NR TXX-3311 81-05 PIPE SPOOL FLANGE MATERIAL NR TXX-3356 81-06 SEISMIC INSTRUMENT TUBING SUPPORT R
TXX-3358 81-07 ORIFICE PLATES SUPPLIED OUTSIDE OF TOLERANCE R
TXX-3527 81-08 UNDERSPECIFIED AFW VALVES NR TXX-3487 1981: 11 TOTAL SDARS--5 REPORTABLE l
s
i<
SDAR INDEX-05-18-87.
t 4
3
. SUBJECT DETERM. STATUS 82-A INSTALLED BORG WARNER VALVES R
TXX-3495 82 MATERIAL PROCURED BY AFC0 STEEL NR TXX-3480 82- 02 DESIGN OF. HORIZONTAL FIRE DAMPERS R*
TXX-3523 82.W RCS WIDE RANGE PRESSURE MEASUREMENTS NR TXX-3571 82-04 PACIFIC PUMP MULTI-VANE DIFFUSER CRACKING NR TXX-3634.
82-05 UNIT 2 SG COMPAR1 MENT WALL-CONCRETE VOID NR TXX-3549 82-06 AUXILIARY SKID UNIT 2 R*
TXX-4095 82-07 DEFECTIVE GOVERNOR DRIVE COUPLINGS R*
TXX-3567 82-08.
DEFECTIVE LIMITORQUE PINION KEYS NR TXX-3580
'82409 SSPS UNDETECTABLE FAILURE.
R*
TXX-3589 82-10 REINSPECTION OF CONDUIT SUPPORTS NR TXX-3583 82-11 OVERTORQUING OF SAFETY RELIEF VALVES NR TXX-3594 82 PIPE-WHIP RESTRAINT WELD INDICATIONS NR TXX-3604-82-13 VOID IN. AUX. BLDG. CONCRETE FLOOR SLABS NR TXX-3589 82-14 REACTOR COOLANT PRESSURIZER SURGE LINE NR
.TXX-3592 82-15' DEFECTIVE PISTON SKIRT CASTINGS R*
TXX-4101 1982: 16 TOTAL SDARS--6 REPORTABLE r
'83-01 BORG-WARNER CHECK VALVE MALFUNCTION R
TXX-3623 83 EMD' VALVE POSITION INDICATION R
TXX-3633 83-03 NON-METALLIC INSULATION NR TXX-3624
.83-04 RADIATION MONITORING. SYSTEM (RM-23)
NR TXX-3635 83-05 VOIDED.
83-06 VENDOR (BAHNSON) INSTALLED HVAC SYSTEM NR
'TXX-4016 83-07 NEW FUEL STORAGE RACKS R
TXX-3677' 83-08 CONTROL VALVE BRACKETS WELDING R*
TXX-3657-83-09 WESTINGHOUSE DS416 REACTOR TRIP SWITCHGEAR R*
W REPORTED 83-10 LETDOWN HEAT EXCHANGER ANCHORS R
TXX-4001 83-11 CCW CLASS V PIPING R*
TXX-3690 83-12 CLASS 1 MATERIAL DEFICIENCIES NR TXX-3691 83-13 STRUT JA>NING DEVICES
- NR TXX-3692 83-14 WESTINGHOUSE NLP PRINTED CIRCUIT CARDS NR TXX-3693 83-15 CABLE TRAY CLAMPS R*
TXX-4005 83-16 WELDED ATTACHMENTS TO PIPING AFTER HYDROSTATIC TESTING NR TXX-4012 83-17 INADEQUATE OVERPRESSURE PROTECTION FOR SPENT FUEL POOL' COOLING HEAT EXCHANGERS CCW RELIEF VALVES R
TXX-4043 83-18 CONTAINMENT BLDG. COOLING R
TXX-4043 83-19 SERVICE WATER SYSTEM VALVES NR TXX-4064 l
83 DEFECTIVE BREAKER SUPPORT BRACKETS R
TXX-4078 83 TRANSMITTER CALIBRATIONS R
TXX-4091 83-22 CHLORINE. DETECTION AND CONTROL ROOM HVAC R
TXX-4098 1983: 21 TOTAL SDARS-13 REPORTABLE i
]
7; LSDAR INDEX-05/18/87-NO.
- SUBJECT DETERM. STATUS-J84. CABLE -TRAY SUPPORT COLUMN (CABLE SPREAD ROOM).NR TXX-4154 84 DIESEL GENERATORS DEFECTIVE PUSH RODS R*
TXX-4108 84-03'. INSTRUMENT TUBING' NR TXX-4110 84-04 FERRORESONANT TRANSFORMERS' R
TXX-4109 84-05
' PROTECTIVE COATINGS ADHESION TESTER NR' TXX-4248 l84 PIPE SUPPORT. DESIGN ERROR NR TXX-4107
.84-07~
' DEFICIENT 4 LUG CRIMPING R
TXX-4135-
.84-08 THERMAL EXPANSION OF CURTAIN FIRE DAMPERS R*
TXX-4257' 84-09
' THRUST BEARING. LUBRICATION NR TXX-4227 84-A INSPECTOR: EYE EXAMINATION RECORDS NR, TXX-4204L
~84-10: ' RODENT DAMAGE:TO CLASS lE CABLES
' NR TXX-4175
'.84-11 UNAUTHORIZED THERMOLAG INSULATION NR TXX-4178 84-12 IMPACT OF.HELB TEMPERATURES ON QUALIFIED EQUIPMENT OUTSIDE-CONTAINMENT P*-
05/28/87.
84-13 FSAR AND TECHNICAL SPECIFICATION CONSISTENCY NR TXX-4213 84-14 ROCKBESTOS CABLE DEFICIENCY
. NR '
~
- 84-15
.INDETERMIN. FLORIDE CONTENT-MANVILLE INSULATION NR TXX-4221
, 84 QUALIFICATION OF MAIN STEAM RELIEF VALVES!
NR TXX-4330
'84-17 FLOODING CONCERNS; R*
'84-18: UNQUALIFIED LIMITORQUE ELECTRIC MOTORS R'
~
84-19 DG MATERIAL ~ DEFECT (VALVE. SPRINGS)
NR.
TXX-4270'84-201 DG. MATERIAL. DEFECT (DELIVERY VALVE HOLDER)-
NR
.TXX-4309 84-21 UT REQUIREMENTS: CLASS 1 PIPING NR
'84-22 CCW SYSTEM DESIGN CONCERN.
NR TXX-4351 84-23' CONAX PENETRATION CONNECTORS.
NR TXX-4375' 84-24 PREOPERATIONAL TESTING ACTIVITIES.
NR TXX-4372 84-25 -AIR-OPERATED DIAPHRAGM VALVES-NR TXX-4318
'84-26 NSSS AUXILIARY RELAY RACXS NR TXX-4317 84-27 VENTILATION EXHAUST DAMPERS-R*:
'84-28 ROSEMOUNT. TRANSMITTERS LEAKAGE
.R*
TXX-4364' 84-29 SAFE SHUTDOWN CAPABILITY.FOR' FIRE-C.R. HVAC
'R*
'84-30 SAFETY CLASS DESIGNATION OF INSTRUMENTS LINES NR TXX-4362 84-31 CONTROL ROOM SEPARATION WALL R*
TXX-4349 84-32 CLASS lE BARTON PRESSURE SWITCHES-PART 21 NR TXX-4363 84-33' RELAYS OUTSIDE SPECIFICATION NR TXX-4376 84 UNIT.2 STRUCTURAL EMBED WELDING NR TXX-4433 1984 35 TOTAL SDARS-10 REPORTABLE. 1 POTENTIAL
[.Q e
SDAR'INDEX-05/18/87
( NO.
SUBJECT DETERM. STATUS 85-01 MAIN CONTROL BOARD CABLE SEPARATION VIOLATIONNR TXX-4411-
- 85-02'.
CABLE' SPREAD ROOM FIRE PROTECTION-3 HR BARRIER NR..
'TXX-4410 L
t85-03' ; STEAM GENERATOR. UPPER LATERAL SUPPORT BEAMS R*-
' 04 CONTAINMENT SPRAY HEADER ISOLATION VALVES R*
TXX-4437 85 DG ENGIUE CONTROL PANEL' AIR FILTER BOWLS:
R TXX-4428 i
85-06 ERF COMPUTER TERMINATION POINTS NR TXX-4447 85-07 ELECTRICAL CABLE BUTT SPLICES:
NR TXX-4490
'85-08 SUMP ISOLATION VALVE MOTORS NR TXX-4481 85-09 VALVE WEIGHT. DISCREPANCIES NR TXX-4473:
85-10 RELIANCE CONTROL BOARD.EQ' DOCUMENTATION t01.
TXX-4523-85-11 INSTRU4ENTATION TUBE FITTING LOCATIONS..
R*
TXX-4454 85-12 AFW PRESSURE' CONTROL-FEED WATER LOW FLOW TO S/G LR*.
TXX-4456 3
85-13 UNDETECTED FAILURE IN SAFETY FEATURES ACT.SYS..
R*
TXX-4457 85-14 UNAUTHORIZED SUPPORT REPAIRS.
R*
-TXX-4465
'85-15 EQUIPMENT. HATCH UNIT'l'UNEVAL. BEARING' FORCES NR TXX-4460 85-16 PHOENIX STEEL-PART 21 WALL THICKNESS QUESTION NR' TXX-4505.
85-17 PART 21-CORE EXIT TEMPERATURE MONITORING ERRORS tGl.
TXX-4485 85-18 SPECIFIC CRITERIA FOR CONCRETE ANCHORS IN THE PROXIMITY OF EMBEDDED. ANGLED NOT DEFINED R*
'85 CONDUIT. SUPPORT SPANS R* ~
08/12/87 85-20 ' WHIP. INTERACTION POSSIBLE-FW BREAK /CS HEADERS R*'
TXX-4516 85-21:
DG POTENTIAL CONTROL PANEL OVERHEATING NR TXX-4493 85-22 LCONTAINMENT ISOLATION VALVES NR TXX-4575 i
85-23 CONTAINMENT SPRAY. PUMP' NR TXX-4563
[.
85-24 TERMINATION OF FIRE DETECTION DETECTORS NR TXX-4533 85-25
. SEQUENCING FIRE PROT. COMPONENTS FOR HVAC/CCW NR TXX-4578 85-26 UPPER LATERAL RESTRAINT EMBEDMENT DESIGN NR TXX-6132
~85-27 SEISMIC GAP LESS THAN DESIGN-CPRT ITEM II.C R*
TXX-4650 85-28 ' PIPE SUPPORTS WELD LENGTH ANALYSES NR TXX-4550 85-29 DESIGN OF ARCHITECTURAL FEATURES R*
TXX-4552 85-30 SWITHGEAR CABINET VENDOR TERMINATION NR TXX-4598 85-31 ELECTRICAL RACEWAY SUPPORT SYSTEM DESIGN R*
06/26/87 85 FEEDWATER REG VALVE SINGLE POWER SUPPLY NR TXX-4600 85-33 NON-SEISMIC TO SEISMIC INTERACTION NR TXX-4577 85-34' CONDUIT SUPPORT SYSTEM R*
02/01/88 85-35 CABLE TRAY HANGER DESIGN R*.
07/06/87 85-36l TRAIN 'C' CONDUIT SUPPORTS Pt 07/08/87 85-37 QUALIFICATION / CERTIFICATION OF INSPECTORS R*
TXX-4569 85-38 CONTROL BOARD SEPARATION MATERIAL NR TXX-4715 4
85-39 EQUIPMENT CONDUIT INTERFACE P*
06/22/87 85-40 FLUX MAPPING SEAL TABLE RESTRAINTS R*
TXX-4633 85-41 COMPUTER PROGRAM ERRORS NR TXX-4716 42 CONDUIT LOADING BY THERMOLAG R*
07/31/87 85-43 QUALIFICATION OF PLANT TEMPERATURES R*
TXX-4659 85-44 QUALIFICATION OF 480V SWITCHGEAR NR TXX-4918 85-45 REBAR DAMAGE NR TXX-4761
~85-46 DAMAGE-STUDY EVAL. OF W-ANALYZED PIPING R*
TXX-4658 85-47 USE OF AMPTECTOR' IIA RELAYS NR TXX-4682 85 48 DG INTAKE SILENCER INTERNALS NR TXX-4683 85-49. MAIN STEAM LINE FLUID TRANSIENT R*
TXX-4656 85-50 CABLE TRAY TEE FITTINGS P*
06/15/87 85-51 APPLICATION OF LOW SULPHUR CONTENT A588 STEEL NR TXX-4722 85-52' CABLE TRAY RANGER REVERIFICATION PROGRAM R*
TXX-4657 85-53 ADEQUACY OF SEISMIC CATEGORY II PLATFORMS P*
06/19/86 85 SEISMIC QUALIFICATION OF HVAC SUPPORTS R*
05/22/87 1985: $4 TOTAL SDARS--23 REPORTABLE 4 POTENTIAL
SDAR INDEX-05/18/87
_NO.
SUBJECT.
o DETERM. STATUS 86-01 WEIDMULLER TERMINAL BLOCKS NR TXX-4712 86-02 TURBINE-DRIVEN AFW PUMP PERFORMANCE NR TXX-4723 86-03 CLASS lE INSTRUMENT TECHNIQUES R*
TXX-4692 86-04 HICHMOND INSERTS TO EMBEDDED PLATE R*
08/01/87 86-05 SAFETY CLASS PIPE SUPPORTS MOUNTED TO NSS EMBEDDED PLATE NR TXX-4776 86-06 PUMP AND DRIVER DOWELING NR TXX-4704 86-07 SERVICE WATER SYSTEM LEAKAGE R*
07/15/87 86-08 SUPER HILTI-KWIK BOLT INSTALLATIONS P*
07/22/87 86-09 DG INLET / EXHAUST VALVE SPRINGS NR TXX-4724 86-10 ELECTRICAL PENETRATION ASSEMBLIES R*
07/10/87 86-11 PUMP IMPELLER LINEAR INDICATIONS R*
09/04/87 86-12 Com0DITY INSTALLATIONS AT SEC. WALL DESIGN GAP P*
06/03/87 86-13 JET IMPINGEMENT LOAD REVIEW P*
06/24/87
,86-14 6.9KV SWITCHGEAR BREAKER WELD FAILURE R
TXX-6341 86-15 DIESEL GENERATOR CONTROLS NR TXX-6002 86-16 FIRE EFFECTS ON INSTRUMENT TUBING R
TXX-6230 86-17 MINIMUM CONCRETE COVERAGE NR TXX-6180 86-18 SAFETY CHILLED WATER CHILLER UNITS R*
12/18/87 86-19 INSTRUMENTATION INSTALLATIONS R*
07/31/87 86-20 WFI NUCLEAR PRODUCTS NR TXX-4870 86-21 QUALIFICATION OF RAYCHEM KIT NPKS-02-01 NR TXX-5046 86-22 SG'S AREA FAN COOLERS P*
05/20/87 86-23 P-10 PERMISSIVE R*
TXX-4835 86-24 SPACE AND MOTOR HEATERS NR TXX-6083 86-25 BREAKER / FUSE COORDINATION NR TXX-4964 86-26 CONTAINMENT SPRAY SYSTEM PIPING NR TXX-4799 86-27 CCW RADIATION DETECTORS NR TXX-4800 86-28 SERVICE WATER SYSTEM DISCHARGE.
NR TXX-4801 86-29 ACCEPTANCE TEST OF AIR OPERATED VALVES NR TXX-4802 86-30 UPS INVERTERS NR TXX-4965 86-31 DG LUBE OIL SUMP TANK FOOT VALVES R*
07/13/87 86-32 THRU-WALL EBMEDDED CONDUIT SLEEVES P*
10/30/87 86 33 STIFFNESS VALUES FOR CLASS 1 STRESS ANALYSIS R
TXX-6025 86-34 DEFECTIVE FIRE STOP INSTALLATIONS NR TXX-4966 86-35 MOTOR OPERATORS FOR MANUAL VALVES R*
08/04/87 86-36 LARGE BORE PIPING AND SUPPORTS R*
06/12/87 86-37 CLASS 1E BATTERY CHARGER COMPONENTS P*
05/08/87
)
86-38 TRACEABILITY OF 1E PIGTAIL EXTENSIONS P*
05/21/87
)
86-39 CABLE TRAY "C-TYPE" CLAMP SHIM DIMENSIONS NR TXX-6403 i
86-40 APPLICATION OF NON-QUALIFIED AGASTAT RELAYS R*
05/29/87 BG -41 SMALL LOCA MODE 4 OPERATION P*
06/12/87 86-42 I&C CABINET POWER SUPPLY BREAKERS NR TXX-4993 86-43 ITT BARTON MODEL 580 SERIES SWITCHES NR TXX-5047 86-44 WELDED ATTACHMENTS TO EMBEDDED STRIP PLATES R
TXX-6261 86-45 SEISMIC CATEGORY II SYSTEMS & COMPONENTS P*
09/11/87 86-46 GOULD BATTERY RACKS TRANSVERSE BRACING R
TXX-6427 86-47 FIRE PENETRATION SEAL DESIGN NR TXX-4996 86-48 ADEQUACY OF NONCONFORMANCE DISPOSITIONS P*
07/15/87 86-49 CONAX ADAPTOR MODULES & ELECTRICAL PENETRATIONS R*
06/19/87 86-50 UNISTRUT SPRING NUTS ON INSTRUMENT SUPPORTS R
TXX-6292 86-51 ANCHOR BOLTS SUPPLIED BY HILTI P*
07/08/87 86-52 CABLE TRAY SPLICES / CONNECTIONS R*
08/03/87 86-53 SEISMIC DESIGN OF CONDUIT R*
09/21/87 86-54 ORIGINAL DESIGN OF CONTROL ROOM CEILING R*
07/21/87 86-55 SEISMIC AIR GAP DESIGN ADEQUACY P*
07/08/87 86-56 DG ENGINE CONNECTING ROD ASSEMBLY NR TXX-5025 86-57 REINFORCING STEEL IN UNIT 1 RX CAVITY WALL NR TXX-6028 86-58 DRILLING / CUTTING OF REINFOR. STEEL IN FUEL BLDG NR TXX-6029 pn_ro ermonom wm t i rwwen.-- " - - -
s l
SDAR INDEX-05/18/87 NQ.
SUBJECT DETERM. STATUS 86-60 POLARCRAEEGIRDERANDGIRDERSUPPORTSHIMGAPS P*
06/17/87 I
86-61 POLAR CRANE RESTRAINTS NR TXX-6092 86-62 POLAR CRANE SUPPORT STRUCTURE R*
06/08/87 86-63 PIPE SUPPORT INSTALLATIONS R*.
07/20/87 i
86-64 COATINGS FOR DIESEL FUEL OIL TANKS P*
05/22/87 86-65 UNCONSOLIDATED CONCRETE-UNIT 2 RCB EXT. WALL NR TXX-6330 86-66 NON-CONTACT LAP SPLICES IN AUX. BLDG.
NR TXX-6282 86-67.
PRE-OP VIBRATION TEST CRITERIA R*
10/22/87 86-68 WEATHER PROTECTION FOR CLASS lE COMPONENTS P*
07/24/87 86-69 SUPPORT OF CLASS lE WIRING P*
04/27/87 86-70 ELEV. TEMP. EFFECTS ON INSTRUMENT SUPPTS& TUBING R
TXX-6220 86-71 CABLE PULLING TENSION P*
06/19/87 86-72 SMALL BORE PIPING AND SUPPORTS R*
06/12/87 86-73 NPSI "XRB" REAR BRACKETS R*
06/12/87 86-74 ROCKWELL TERMINAL BOXES NR TXX-6372 86-75 ASCO SOLEN 0ID VALVES IN PISTON AIR ACTUATORS OF PAPC0 DAMPERS NR TXX-6287 86-76 DEVIATIONS IN GOULD BATTERY RACK DIMENSIONS P*
06/05/87 86-77 INSTRUMENT TUBING MINIMUM WALL REQUIREMENTS R
TXX-6228 86-78 STRUCTURAL STEEL BOLTING IN CABLE SPREAD ROOM P*
06/04/87 86-79 OVERTORQUED WESTINGHOUSE AR RELAYS P*
06/22/87 86-80 TORQUE VALUES FOR J.C. WHITE TUBING RESTRAINTS NR TXX-6140 t
86-81 BOP INSTRUMENT SETPOINT ERRORS P*
07/29/87 86-82 CTH WELDS USED TO SPLICE CHANNEL SECTIONS R*
07/17/87 86-83 SBM ON POWER CABLES AND POWER RACEWAYS P*
06/05/87 1986: 83 TOTAL SDARS--31 REPORTABLE, 22 POTENTIAL i
1
- ,L=d.*.
8M
' SUBJECT ;.
DETERM. STATUS c87-01 ' STATIC COMPUTER MODEL FOR SG BLDG P*
07/29/87 87-02 FISHER CONTROL VALVES NR TXX-6316
= 87-03.:
6.9 KV SWITCHGEAR INSTALLATIONS ~
R.
)
[
87-04' PLUG' WELDING ON EMBEDDED CHANNEL P*
06/17/87
-87.-05 EQ OF CABLE TO POST-LOCA HRRM' P*
07/21/87 87;-06 SQ OF AS-BUILT 480 VOLT POWER CENTERS-
.P*
09/04/87
'87-07.
PROCUREMENT OF SPARE PARTS W/0 QUAL.REQTS.-
P*.
02/19/88-
.87-08 CAL. ACCURACY OF PRES.STD. DEAD WEIGHT. TESTER P*
05/20/87 l
g
- 87.* 09 - TERMINAL STUDS.IN PK-2 TEST BLOCKS P*
06/05/87 j 10 CONTROL OF DESIGN MODIFICATIONS P*.
06/10/87
'87-11' DEFECTIVE CIRCUIT' CARDS IN BATTERY CHARGERS P*
06/12/87 i
87-12. CONTAINMENT P/T ANALYSIS' COMPUTER ERROR P*
06/12/87 i
87-13 CLASS 1E SEPARATION VIOLATIONS P*-
06/15/87 i
[
~1987: 13 TOTAL SDARS--
l REPORTABLE, 11 POTENTIAL l-
1
-=M Log # TXX-6381 Fi1e #10110
~
907.3 C
C Ref: 10CFR50.55(e) 1UELECTRIC w un.ni c. cm.ma June 5, 1987 f.w wne i h e Preudent U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555
SUBJECT:
' COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 OPERATION OF SAFEGUARD ACTUATION RESET CIRCUITRY (IEB 77-03)
SDAR:
CP-77-D (FINAL REPORT)
Gentlemen:
Our letter, logged TXX-2481 and dated September 20, 1977, was concerned with Westinghouse AR Relays with latch attachments (Ref. IE Bulletin No.~ 77-02, your letter dated September 12, 1977) and on-line testing of Westinghouse Solid State Protection System (Ref. IE Bulletin No. 77-03, your letter dated September 12,1977). This was a reportable issue per 10CFR50.55(e).
Both bulletins were concerned with Westinghouse AR Relays.
Bulletin IE 77-02 reported a design deficiency for certain types of Westinghouse AR Latching Relays and was a hardware problem.
Bulletin IE 77-03 was a software problem, identifying modifications to on-line test procedures for the Westinghouse Solid State Protection System (WSSPS), which had to be made in order to detect contact closure failure of the Westinghouse AR Relay which would not normally be detectable by any other means.
In our letter to you we inadvertently concluded that the second bulletin (IE 77-03) covered changes which had to be made to operating procedures only if the relays covered by the first bulletin (IE 77-02) were utilized at the pl ant. We therefore advised that "since we will be receiving only corrected equipment, this augmented testing is not applicable".
After further review, we have concluded that the modification to the procedure is necessary to periodically completely test the safeguard actuation reset circuitry.
Should any part of the circuit (including any relay) being tested by the modification fail, safety injection could not normally be reset and/or blocked to allow initiation of recirculation.
400 North Olise Street I.11 til Dallas. Texas 75201 l
l s
June 5, 1987 Page 2 of'2 Although this issue is not being tracked as an open item by the Commission (See NRC IR 50-445/84-22,50-446/84-07), we wish to inform you that the corrective action is being implemented. Operating procedures (ICI-4013A &
ICI-4014A Unit 1 Train "A" and Train "B", respectively) have been revised and are in review prior to being issued.
The Unit 1 procedures will be issued by September 30, 1987.
After Unit I procedures are issued, similar modifications will be incorporated for Unit 2 procedures and they will be issued by June 30, 1988.
4 Very truly yours, (7?%
/
-W. G. Counsil I
JCH/mlh c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3 copies) i I
4 J
t Log # TXX-6482 I
==
File #-10110 i
l I
903.11 i
=.
.C Ref. #'100FR50.55(e).
1 q
~
nlELECTRIC May.28, 1987-wn.,n c. cui Eterunve Vice />rsidem ^
('
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
- I DOCKETS NOS. 50-445 AND 50-446 QUALIFIED EQUIPMENT OUTSIDE CONTAINMENT SDAR:
CP-84-12-(INTERIM REPORT)
Gentlemen:
On June 4,1984, we verbally notified your Mr. D. Hunnicutt of a deficiency involving the environmental qualification of equipment outside~ containment for high energy line breaks (HELB). This is an interim report of a potentially reportable. item under provisions of 10CFR50.55(e).
Our latest interim report, logged TXX-6360, was submitted on March 27, 1987.
1 The analysis of this issue is continuing. We will submit our next report no later than July-28,1987.
Very truly yours, W. G. Counsil
/ l L
By:
G. S. Keeley e-J Manager, Nuclear 1.icVnsing BSD/amb c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3)
O 4m North (he Mreet i H Wf Dab h as"?ni
l i
M Log # TXX-6421 p
FE File # 10110 l
903.11
=
=
Ref: 10CFR50.55(e)
TUELECTRIC m.
c.o,. (
April 30, 1987 Derutne ine Preudem U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk i
Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET.NOS. 50-445 AND 50-446 CONDUIT SUPPORT SPANS SDAR:
CP-85-19 (INTERIM REPORT)
Gentlemen:
On October 21, 1986, regarding the use of a nonrepresentative sample in determining generic typical conduit span calculations which we deemed reportable under the provisions of 10CFR50.55(e).
This is an interim report to provide additional technical information involving the deficiency and to status corrective actions implemented to date.
As previously noted in our letter logged TXX-6046, the deficiency involves previous design requirements and documents for safety related conduit and conduit supports for Units 1 and 2.
Specifically, the deficiency concerns the yield stress (Fy) used to justify the span calculation reflected in the previous design.
During the Independent Assessment Program (IAP) conducted by Cygna, two major comments were noted on the derivation of the conduit allowable stress.
a) The allowable stress values vary with the nominal size of the conduit.
Using vendor test data (three to four tests for. each size) the design specified the lowest yield stress or an imposed minimum yield stress value (33KSI) to obtain allowables for each conduit size.
The issue observed by Cygna questioned the different allowable stress for each size and the lack of justification for the imposed minimum yield stress.
b) The conduit used in CPSES installations is supplied by Triangle PWC and fabricated in accordance with ANSI C80.1, which contains no request for
-material conformance.
Considering the supplier only processes and does not manufacture the material used in the conduits, and no certification is provided with the conduit as a rule, the test data represents only a general sample of conduit.
i 4
400 North Olive Street L B 81 Dallas, Texas 75201
)
TXX-6421 April 30,1987 a
Page.2~
1 These observations resulted in a concern that, in the event of a postulated seismic event, Category 1E conduits, which did not meet the required design criteria for yield stress, could have failed in a manner adverse to plant safety.
Our current engineering effort is intended to assure the final basis for the conduit " yield stress" meets the criteria specified in the FSAR.
An expanded sample of available material test reports (MTR) for all conduit sizes has been reviewed.
This review has indicated the number of MTRs used to establish the yield stress in the original design was too limited, and that the use of the lower test yield stress per conduit size is inappropriate.
Based upon the completed evaluation of an expanded MTR sample, the design documents, 2323-S1-0910 for Unit 1, and 2323-S2-0910 for Unit 2, are currently scheduled for revision by July 31, 1987.
At that point, additional corrective measures will be defined if required.
'The conduit and conduit support systems are currently undergoing a 100Y. design verification.
In addition to this deficiency, the cumulative effects of all CPRT design, construction and quality issues related to these systems are being evaluated per SDAR CP-85-34.
We anticipate submitting our next report on this issue no later than August 12, 1987.
Very truly yours, W. G.
unsil
/
s A b-By:
~
- 3 f 1
G. S. Keeley Manager, Nuclear Lice sing JCH/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV
r-f V
j
,y-j y
F====
Log # TXX-6438
-- E File # 10110 903.10'
_=
=
Ref: 10CFR50.55(e)
TUELECTRIC William G. Counhil tm.,re,er,w""
j May 12, 1987 U. 'S; Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 EQUIPMENT CONDUIT INTERFACE SDAR:
CP-85-39 (INTERIM REPORT)
Gentlemen:
On September 18,1986, we verbally notified your Mr. T. F. Westerman of a deficiency involving conduit which was not installed in accordance with design documents. This is an interim report of a potentially reportable item under the provision of 10CFR50.55(e).
Our most recent interim report was logged TXX-6278, dated February 12, 1987.
As previously reported, the program to evaluate this condition consists of the following three issues:
1)
Cable slack adequacy.
2)-
Conduit-to-equipment interfaces outside the vendor approved area.
3)1 Conduit orientation within the vendor approved entry area
'(completed per Interim Report logged TXX-6165).
The cable siack adequacy issue utilized a sampling approach to evaluate as-built
' cable slack. As previously advised,' evaluation of the sample was scheduled to be l
' completed and reported by February 27, 1987. The preliminary report was issued February 23, 1987, and is under Third Party Review. The final report is scheduled to be' issued May 15, 1987..
Field verification activities have been completed on 95% of the class IE equipment in Unit 1 for the conduit-to-equipment interface issue and has identified approximately 100 examples in which conduit enters equipment outside the vendor approved entry area.
The final analysis of each configuration with respect to seismic qualification is scheduled to be issued by June 1,1987.
400 North Olive Street L H. 81 Dallas, Texas 75201 b
~
K.
, May 12,L1987 Page 2,.of 2%
The completion of the above activities.is. required before the evaluation of this issue'and the effects of this issue ~on the safety of plant operations can be completed.
We will submit our next report'on this issue no later than June 22, 1987.
Very truly yours, Yk W. G. Co nsil
[./
By:
G. S. Keeley &
/-
Manager,NuclearLpensing RSB/mlh c - Mr_. R. D. Martin, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 4
J g
-s 9
E Log # TXX-6413 File # 10110 C
908.3 C
Ref: 10CFR50.55(e) tilELECTRIC
..,,,, c, c, April 28, 1987 Emuun She Prrudent U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
. COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 CABLE TRAY TEE FITTINGS SDAR:
CP-85-50 (INTERIM REPORT)
-Gentiemen:
On December 6,1985, we verbally notified your Mr. H. S. Phillips of a deficiency involving the use of improper welding processes in the manufacturing of cable tray tee fittings which could affect the equipment qualification.
This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e).
Our latest interim report, logged TXX-6151, was submitted December 15, 1986.
As previously stated, a testing program is being conducted to establish and demonstrate minimum weld requirem'ents and to determine whether weld conditions of installed components are acceptable. The test program is scheduled to be completed on May 15, 1987.
We will submit our next report on or before June 15, 1987.
Very truly yours, W. G. Counsil By:
~
G. S~ Keeley -
L-Manager, Nuclear %nsing RSB/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 400 North Olive Street L B. 81 Dallas, lexas 75201 i'
s.
7 i
=
Log # TXX-6460 P9 File # 10110 L
903.7 4
r C
Ref: # 10CFR50.55(e) 1UELECTRIC Milham G. Counnit ununve ser neuac,,o Hay 22, 1987 U. S. Nuclear _ Regulatory Commission Attn: _ Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC' STATION-(CPSES)
DOCKET NOS. 50-445 AND~50-446
~ EISMIC QUALIFICATION OF HVAC SUPPORTS S
SDAR: CP-85-54 (FINAL' REPORT)
Gentlemen:
On September 26, 1986, we notified you by letter TXX-5043 that a deficiency involving discrepancies in supplier provided HVAC support documents was i
reportable under the provisions of 10CFR50.55(e).
Our last interim report logged was TXX-6106 dated November 20, 1986.
This is a final report to status the following corrective measures:
1 1.
The Project organization and responsibilities for Units 1 and 2 HVAC 1
activities have been realigned.- Engineering and QC activities previously performed by Bahnson have been assumed by Ebasco and TV Electric, respectively.
2.
The field verification effort has been expanded to encompass all Unit 1 and 2 Seismic Category I duct and duct supports.
]
Ebasco's progress to date, for this effort, is as follows:
Collection of existing documents is ' complete for Units 1 and 2.
l O
l 0
The as-built inspection of duct supports in Unit I and Common has been completed with the exception of areas which will be verified.
The as-built inspection of ductwork.in Unit 1 and Common is currently in progress and is being performed in accordance with TV Electric Procedure CPE-FVM-CS-029, " Procedure for Seismic HVAC Duct and Duct Hangers As-Built Verification in Unit I and Common Areas."
3.
Engineering evaluations are being performed to ensure compliance with FSAR commitments.
4.
Construction re-work was initiated on May 4, 1987, as deemed necessary by the engineering evaluations of field verified information, to ensure compliance with FSAR criteria for the affected supports.
400 North Okve Street L B 81 Dallas. Texas 75201 l
E V,
TXX-6460f May 22,-1987 Page 2 of 2-L5.- QC inspections of the' Seismic Category 1 duct and duct supports.will be performed after an engineering review of the construction re-work.
- 6. : An as' built /and design, verification' effort, and QC inspection will be performed on Seismic Category I duct and duct supportstin Unit 2.
This-program will be_ performed in the same manner as the one ' utilized in Unit-1.L
' Unit 1 activities'are scheduled for completion by March, 1988.
Unit 2 l activities are scheduled for completion in July,1988.
Very truly yours, W. G..Counsil RWH/gj
'=
c - Mr..R. D.-Martin - Region IV CPSES Resident Inspector - 3 copies 6
)
i P9
. Log # TXX-6423 1
=
' File # 10110 r
903.8' H# ELECTRIC.
Ref # 10CFR50.55(e)
WHumm G. Coumou Lwutive Vke Presusent May 1, 1987 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
)
1
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
-DOCKET NOS. 50-445 AND 50-446 PR0XIMITY PLACEMENT - RICHMOND INSERTS TO EMBEDDED-1 PLATES AND WELDED ATTACHMENTS TO EMBEDDED PLATES
-1 s
- SDAR: CP-86-04 (INTERIM REPORT)
Gentlemen:
On February 13, 1987, we notified you by our letter logged TXX-6260, of a a
deficiency involving the placement of Richmond inserts in the proximity of loaded embedded plates which could compromise the' integrity-of adjacent shear cones.
This is.an interim report of a deficiency which-is reportable under the provisions of 10CFR50.55(e). By our. telephone request. of April 29 1987,
. received by.your Mr. Cliff Hale, the due date for this report was extended until May-13, 1987.
Our corrective act' ions previously specified for Unit 1 are continuing as planned.
Based on the current project schedule, it is anticipated that the Unit I corrective actions will be completed by December 31, 1987. New work for Unit 2 is being done in accordance with the revised design specification and the related construction / quality control procedures. The Unit 2 field verification of.. existing installations will be conducted following ~the completion of the Unit I verification.
We anticipate submitting our next report no later than August 1,.1987.
Very truly yours, W. G. Counsil By:
G. S. Keeley
/
Manager, Nuclear Lit /nsing c - Mr. E. H. Johnson, Region IV Mr.'D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV
^
400 North Olive Street L B. E l Dallas, Texas 75201
Log # TXX-6467 File # 10110
'908.1 C
Ref #10CFR50.55(e)
NELECTRIC
%illiam G. (ounsit a,, ou,, n, e,,w,m June 1, 1987 U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555-
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 ELECTRICAL PENETRATION ASSEMBLIES SDAR:
CP-86-10 (INTERIM REPORT)
Gentlemen:
On April 24, 1986, we notified you by our letter logged TXX-4774 of a
' deficiency involving the electrical penetration assemblies (EPA's) supplied by Bunker Ramo which we deemed reportable under the provisions of 10CFR50.55(e).
i This is a. follow-up interim report submitted to. status corrective action implemented to date.
Our last interim report, logged TXX-6357, was submitted March 27, 1987.
{
As noted in previous reports, the EPA modules supplied by Bunker Ramo are I
being replaced by modules supplied by Conax.
In our letter logged TXX-6262, dated February 27, 1987,- concerning Enforcement Action (EA) 86-09, we stated that future schedule changes for corrective actions relating to the replacement of these modules would be updated via SDAR-CP-86-10.
Rework is in progress and is currently scheduled for completion by July 30, 1987 for Unit 1 and November 30, 1987 for Unit 2.
We expect to submit our next report by December 18, 1987.
Very truly yours, Mkt W. G. Counsil BSD/dl c - Mr. R. D. Martin - Region IV Resident Inspectors, CPSES (3) i I
400 North Olne Street
- 1. H N I Dallas. Texas 75201 j
g,R y: f f
- 7 m n y
+
1.
g I
=
Log # TXX-6474 File # 10110 C.
903.7
~~
C Ref:
' TUELECTRIC W h m C.Counsa c=w,nc,rmw.
June 3, 1987 U. S. Nuclear Regulatory Commission
' ATTN: Document Control Desk Washington, DC 20555 3
SUBJECT:
COMANCHEPEAKSTEAMELECTRICSTATION(CPSES)
DOCKET NOS. 50-445 AND 50-446 COMMODITY INSTALLATIONS AT SECONDARY WALL DESIGN GAPS' SDAR:
CP-86-12 (INTERIM REPORT)
Gentlemen:
On February 27, 1986,- we verbally notified your Mr, T. F. Westerman of.a deficiency involving attachments at, or in close proximity to, design gaps between secondary walls and adjacent walls / floors. This is an interim report of.af potentially reportable item under the provision of 10CFR50.55(e). - Our last interim report logged was TXX-6362, dated March 27, 1987.
The methodology for. evaluation of commodities attached to secondary and adjacent. structures or walls has been developed and incorporated in Design Bases Document (DBD CS-19) " Building'and Secondary Wall Displacements".
Evaluation of the commodities bridging gaps between secondary walls and adjacent structures is now scheduled to be completed by the end of June,1987,
?
for Unit 1 and common areas. Commodities that are deemed unacceptable by evaluation will be reworked as required. Unit 2 will be scheduled after completion of Unit 1.
A Commodity Clearance Program is being developed for all seismically supported commodities and will include those installed in close proximity to secondary walls. Verification Method CPE-FVM-CS-068 " Commodity Clearance" has been drafted and is in interdisciplinary review.
In addition, Specification CPE-S-1021 is being prepared to insure that no future installations violate specific clearance criteria.
The evaluation of commodity clearances is expected to be completed by November, 1987.
The methodology for determining the structural adequacy of secondary wall designs which did not previously consider support loadings has been established and Field Verif-ication Method (FVH CS-058) was issued on May 8, 1987. The evaluation of secondary walls, considering commodity support loading, is scheduled for completion by July 30, 1987.
400 North Olwe Street t..H 81 Ddlas. Texas 75201
_ TXX-6474' June 3, 1987 Page 2.of 2' o
Following a full analysis of these issues, an evaluation will be made regarding fhe effects on the safety of plant operations had these conditions gone uncorrected.
We will submit our next report by December 18, 1987.
Very truly yours, ndAb W. G. Counsil
/
By:
G. S. Keeley &
/-
Manager, Nuclear y ensing JCH/mlh c - Mr. R. D. Martin, Region IV Resident Inspectors,'CPSES (3 copies) j 1
M Log # TXX-6459 FE File # 10110 909.4
.=
=
Ref # 10CFR50.55(e)
TUELECTRIC
%Esm G (ounsa L m wng i u r Perudem U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 SAFEGUARDS AREA FAN COOLERS SDAR:
CP-86-22 (INTERIM REPORT),s Gentlemen:
On April 2, 1986, we verbally notified your Mr. T. F. Westerman of a deficiency involving the capacity of the present fan cooler design in the Safeguards area to adequately handle the heat load in several rooms.
This is an interim report of a potentially reportable item under the provision of 10CFR50.55(e). We have submitted interim reports logged TXX-4780, TXX-4894, TXX-5008, and TXX-6158 dated April 30, 1986, July 3, 1986, September 12, 1986, and December 12, 1986, respectively.
During a review of Unit 2 preoperational test procedures, a comparison of fan cooler specifications and drawings identified cooling capacities below the calculated heat loads in the design basis calculation.
Test performance data for Unit I coolers may have been accepted at values below both the design heat loads and equipment capacities.
In order to assess the impact of this condition on the safety of plant operations, an engineering evaluation was undertaken to determine if failure of equipment cooling capacities to meet required design criteria could result in an inability of associated safety-related systems to perform as required following LOCA with loss of offsite power.
The completed calculations for Unit I have determined that the emergency fan coolers (served by the safety chilled water system) are adequately sized to maintain the room temperatures, within the FSAR requirements, following a LOCA coincidental with loss of offsite power.
However, we have determined that the chilled water flow rates require rebalancing.
400 North Olne Street L B. 81 Dalian Texas 75201
lg Q s,
I,
TXX-6459' May 20, 1987
. Page 2 'of '2 We are continuing with our evaluation which' includes' revising the design 4
documents as indicated above.
We. expect to; submit our next report by' July 8, 1987.
Very truly yours, E!'I, [c,, ;_ lI
W.' G. Counsil l
~~
By:
- 5. S. Marshall Supervisor, Generic Licensing
. GLB/gj c - Mr. R. D. Martin - Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI.- Re~gion IV k
Pa===
Log # TXX-6446 E ""-
File # 10110 903.9
.r r
Ref: 10CFR50.55(e) 1UELECTRIC
%&lham G. Counsil e m um, s o, n,w,n, May 15, 1987 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 MOTOR OPERATORS FOR MANUAL VALVES SDAR:
CP-86-35 (INTERIM REPORT)
Gentlemen:
On June 2, 1986, we verbally notified you of a reportable' item involving the-motor operators for several valves in the main steam system (see TXX-4840).
Our most recent report was submitted September 12, 1986.
This report is i
submitted to status corrective actions implemented to date.
In our previous report we indicated that design activities were scheduled to be completed by April 1987.
These activities are now scheduled to be completed this month.
These design changes are currently being developed under DM 86-106
" Replacement of Main Steam System to Steam Generator Blow-down System HELB Isolation Valves" (Unit 1) and PPF-2-1029 " Replacement of Main Steam System to Steam Generator Blow-down System HELB Isolation Valves" (Unit 2) to modify the existing design.
Engineering has completed procedure PP-074 " Engineering and 3
Design Requirements for ASME XI Repairs and Replacement / Modifications".
These requirements are being implemented in the design documents included in DM 86-106 and PPF.-2-1029 above for Units 1 and 2.
The redesign requires the modified manually operated valves be restored to their original condition and additional air-operated valves be installed to provide for the double HELB isolation capabilities described in FSAR Section 7.6.12.2.
A purchase order has been placed for eight air-operated valves for Unit 1 and 2.
We anticipate submitting our next report on this issue no later than August 4, 1987.
j Very truly yours, t
W. G. Counsil JDS/mlh c - Mr. R. D. Martin, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV j
400 North Okve Street L B 81 Dallas, Texas 75201
y P_
Log # TXX-6437
'4 y9 File # 10110
?
908.3 r
r Ref: 10CFR50.55(e) 1UELECTRIC
$5Nd[bu, May 8, 1987 U. S.' Nuclear Regulatory Commission
, ATTN:- Document Control Desk Wa'shington,.D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 CLASS IE BATTERY CHARGER COMPONENTS t
.SDAR:
CP-86-37 (INTERIM REPORT)
Gentlemen:
.On May 14,.1986, we verbally notified your Mr. I. Barnes of a deficiency involving
' firing board assemblies and amplifier boards used in Class 1E battery chargers that do not conform to' vendor assembly drawings.
This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Our most recent interim report was logged TXX-6276, dated February 11, 1987.
Our review of the procurement processes and adequacy of the affected hardware is continuing. We are unable to complete our assessment of this issue for plant safety significance until the review is completed.
We expect to submit our next report by June 26, 1987.
w by Very trul yours, m
W. G.
ounsil By:
N!
G. S. Keeley Manager, Nuclear Li sing WJH/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 400 North Olne Street
- 1. H. 81 Dallas, Texas 75201
r Log # TXX-6455 P """:
File # 10110 L
908.3
=
=
Ref _ # 10CFR50.55(e) 1UELECTRIC
%illum G, Counsil i.m mn, s ua,,,,sen, May 20, 1987 U. S. Nuclear ' Regulatory Commiss' ion Attn:
Document Control Desk Washington, D.C. _20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 s
TRACEABILITY OF IE PIGTAIL. EXTENSIONS SDAR:
CP-86-38 (INTERIM REPORT)
Gentlemen:
On May 27, 1986, we verbally notified your Mr. T. F. Westerman of a deficiency involving the failure of project procedures to adequately address the installation and inspection criteria for cable runs used as device pigtail extensions. -This is an. interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Our latest interim report, logged TXX-6256 l
was submitted on February 4, 1987.
- The Unit 1 field verification effort to obtain cable and splice data is i
co;aplete.
The evaluation of the as-built data for Unit I and our assessment of safety significance is scheduled to be completed by September 10, 1987.
A field verification program for Unit 2 will be performed in the same manner as the program implemented for Unit 1.
This effort is scheduled to begin in July, 1987, and be completed in March, 1988.
Our next report on this issue _will be submitted no later than October 5, 1987.
Very truly yours, awaa W. G.
ounsil By:
G. S. Keeley-C '
Manager,NuclearLNensing RSB/mlh-c - Mr. R. D. Martin - Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 400 North Othe Street Lh 81 Dalla 3. Texas 75201
,iL e
r e
M Log # TXX-6403
==
File # 10010 d
903.9
=
=
Ref: 10CFR50.55(e) nlELECTRIC April 23, 1987 Wmiam G. Coumit becutwe %ce Pressdent U. S. Nuclear Regulatory Commission I
ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM. ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 CABLE TRAY "C-TYPE" CLAMP SHIM DIMENSIONS SDAR:
CP-86-39 (FINAL REPORT)
Gentlemen:
On May 27, 1986, we verbally notified your Mr. T. F. Westerman of a deficiency involving dimensional tolerances of the shims used in. conjunction with "C-Type" clamps on cable trays.
Our last interim report, logged TXX-6123, was submitted November 26, 1986.
We requested and received.an extension until April 23, 1987, in providing our response during a telephone conversation with Mr. I. Barnes on April'21, 1987.
The original design requirements specified the shim plates, when installed with "C-Type" clamps, to be flush with the tray web and extend a. maximum of 1/2" beyond the outside of the web. This configuration is required to provide shimming under the full width of the tray flange.
Due to an error in transferring information from the original design change to the erection and inspection drawing, the shim plate tolerance of plus or minus 1/2" from the tray' web was specified allowing the shim plate to be installed up to 1/2" inward from the web underneath the tray flange. As a result, some unacceptable clamp / shim configurations have been installed and accepted, j
A study documented SAG TUG 2.3494, Vol. 1, Book 17 was performed indicating local I
stresses in the tray flange near the "C" clamp, due to the maximum anticipated loading conditions, were less than the allowable plastic limits given in the CPSES FSAR (section 3.8.4.3.3) except for the T. J. Cope 4" x 12" tray section.
A testing program conducted by Corporate Consulting and Development Company (CCL) evaluated the T. J. Cope 4" x 12" tray / clamp configuration. A 4" x 36" tray was also included for testing for worst case loading conditions.
Evaluation of these test results indicates that the potential condition of "C" clamps installed with a shim plate located 1/2" inward from the web is acceptable.
These results have been documented in CCL Test Report No. A-738-87 and SAG TUG 1.8160 2.C.S.
400 Nonh Olive Street LB 81 Dallas, Texas 75201
r--
y
" "# ~~ - TXX-6403' April 23, 1987 PagC2 o'f 2!' '
We have concluded that had this deficiency remained uncorrected, it would have no adverse effect on the. safety of plant operations.
This issue is'not reportable under the provisions of 10CFR50.55(e).
Records supporting this evaluation are available for your inspectors' review at the CPSES site.
Very truly yours, f d d (.5 tubi-W. G. Counsil 45 -
By:
't Jr S. Marshall Supervisor, Generic Licensing RSB/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI Region IV Mr. H. S. Phillips, RI - Region IV
t Log # TXX-6475 L'*
File # 10110
-= = -
903.11
{ ""_";"
Ref # 10CFR50.55(e)
-~
l 1UELECTRIC May 29, 1987-Milliam C. (nunsil Ltreutive l'ur President U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 APPLICATION OF NON-QUALIFIED AGASTAT RELAYS SDAR: CP-86-40 (INTERIM REPORT)
Gentlemen:
On October 10, 1986, we notified you by our letter logged TXX-6017 of a deficiency involving the procurement and use of non-qualified Agastat relays in several Class IE applications which we deemed reportable under the provisions of 10CFR50.55(e). This is an interim report submitted to status corrective action implemented to date.
Our most recent interim report was logged TXX-6313, dated March 2, 1987.
We have completed the identification of non-qualified relays affected by this deficiency and plan to implement the following corrective actions:
1.
All future Agastat 7000 series relays will be purchased with an "E" prefix for use in Class IE application, and will be segregated in the warehouse from other relays without an "E" prefix.
2.
7000 series Agastat relays without an "E" prefix, procured in the field with Q Code-C (commercial) for IE application and stored in the warehouse, are to be replaced with "E" prefixed relays.
Agastat 7000 series relays without the "E" prefix will be re-classified for use in non-1E applications.
3.
7000 series Agastat relays without an "E" prefix, procured in the field with Q Code-C and installed in safety related equipment, are 1
to be replaced with "E" prefixed relays.
The relays without an "E" prefix will be returned to the warehouse for use in non-1E applications.
4.
7000 series Agastat relays without an "E" prefix will be listed in
" Nuclear Operations Defective Items List" (N0DIL) to prevent the future use of the relay without an "E" prefix for IE application.
i 1
1 400 North Olne Street L B. 81 Dallas, Texas 75201 J
-May 29, 1987-Page 2 of 2 The replacement of the non-qualified relays with qualified relays will ensure that the-equipment installation conforms to the equipment qualification specifications.
These actions will be accomplished in accordance with Nonconformance Report (NCR) 85-0004,- Revision 3, and are scheduled for completion by September 18, 1987.
t Also, during our review, we have determined that Westinghouse notified TV Electric at the time the equipment was shipped (as noted in the document package) that the equipment had been furnished with unqualified relays which would require subsequent replacement. The equipment was installed with the unqualified relays.
Programs are currently established to ensure similar conditions (i.e., installed equipment with unqualified componsnts or equipment purchased in the future) are identified and dispositioned, l
We will submit our next report no later than October 9,1987.
Very truly yours, lf W. G. Counsil WJH/gj c - Mr. R. D. Martin - Region IV CPSES Resident Inspector - 3 copies W
e-M Log # TXX-6427 FE File # 10110 908.3
.=
=
Ref: 10CFR50.55(e) 1UELECTRIC y,,,,,_,
April 30, 1987 bewwr Vue Hessdem U. S. Nuclear Regulatory Commission ATIN: Document Control Desk i
Washington,- DC 20555 i
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 G0ULD BATTERY RACKS TRANSVERSE BRACING SDAR: CP-86-46 (FIN /.' REPORT)
Gentlemen:
1 On October 21, 1986, we notified you by our letter, logged TXX-6052, of a deficiency involving a design change to vendor supplied equipment being made without formal evaluation by the vendor, which we deemed reportable under the provisions of 10CFR50.55(e). This is a final report submitted to status i
corrective action implemented to date.
In our most recent transmittal, our corrective actions were noted as follows:
a)
The design change which authorized the one (1) bolt connection has been voided and the original design requirements reestablished, thus reverting back to the original design of the two (2) bolt connection.
i b)
TUGC0 Nuclear Engineering Procedure TNE-DC-8-1, " Engineering Review and Design' Verification of Field Design Changes," (currently Rev.10) has been revised and now provides more detailed instructions and establishes a verifiable approval record of all design changes including calculations.
Completion of the ass 5ciated rework concerning both Units 1 and 2 is scheduled for August 31, 1987.
Very tr ly yours, W. G. Counsil l
.k< s By:
G. S. Keeley Manager, NuclelF pc,e/-
nsing RWH/mlh c - Mr. E. H. Johnson, Region IV
\\J i
Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV i
I 400 North Ohve Sirce LB. 81 Dallas, Tetas 75201
1 s,
c
=7 Log # TXX-6415
- 9 File # 10110 903.11
=
=
Ref: 10CFR50.55(e) 1UELECTRIC.
April 29, 1987 m.m c. coumu larsunve ii<r heudent U. S. Nuclear Regulatory Commission ATTN:
Document-Control Desk-l Washington, DC' 20555 i
I
SUBJECT:
~ COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 CABLE TRAY SPLICES / CONNECTIONS SDAR:
CP-86-52 (INTERIM REPORT)
Gentlemen:
l On July 25, 1986, we verbally notified your Mr. I. Barnes of a deficiency involving snlice/ connections used for cable trays in Units 1 and 2.
We have-concluded that this deficiency is reportable under the provisions of 10CFR50.55(e) and the required information is as follows.
We requested and received an extension until May 8, 1987, in providing our response'in a telephone conversation with Mr. I. Barnes on April 28, 1987.
Descriotion of the Problem During as-built activities on the cable tray hangers for Unit 2, a' deficiency involving incorrect installation of splice / connections was identified.
Specifically, the splice / connections used for some tray types and sizes were not
' installed in accordance with specified splice patterns as defined by the cable tray manufacturer or approved CPSES design or design change documents. The QC Inspection requirements delineated by procedure QI-QP-11.3-24, " Class IE Cable Tray Raceway Inspections," did not include verification of the splice / connector configurations or dimensions.
Engineering evaluation of the as-built information for both units has indicated that a potential failure could occur in "Z" type splices. These splices have been categorized as potential failures, because these connections would be expected to fail when subjected to seismic loading conditions.
Splice configurations for both units which require further engineering evaluation have been selected for testing.
The test specification has been completed and the test is expected to begin by May 4, 1987.
Upon completion of testing, an J
engineering evaluation of the results will be performed and additional recommendations made for corrective actions if required.
Safety Imolication i
Failure of cable trays to perform as intended by engineering design could result in an inability of associated safety-related electrical systems to perform as required for safe operation / shutdown during a seismic event.
400 North Olive Street L B. 81 Dallas. lexas 75201
g
. April'29, 1987
- Page 2 of 2 Corrective Actions'-
1
- The following measures have been or will be implemented to prevent the misuse of splice plates in Units 1 and 2.
I 1.
Design Change Authorizations (DCA's) have been issued to provide criteria per specifications ES-100, " Electrical Erection," via DCA 31431 and ES-19, " Cable Tray Specifications," via DCA '34029.
Previously i
generated splice ' plate DCA's have been voided to prevent use, f
1 2.
Following the issuance of these DCA's, electrical construction procedure ECP-10, " Cable Tray and Hanger Installation Unit 1," and ECP-10A, " Cable
]
Tray and Hanger Installation Unit 2," will be revised to indicate acceptable splice plate configurations.
3.
NCR's will be generated to identify specific occurrences of the use of "Z" type splices and document specific measures.
4.
Rework on future installations will be accomplished using approved procedures and specific DCA's against unique tray sections.
5.
QC Inspections Procedure QI-QP-11.3-24, " Class 1E Cable Tray Raceway Inspections," will be revised to include inspection of splice plate configuration and dimension criteria.
Based on the projected schedule, our next report will be issued no later than August 3, 1987.
4 Very truly yours, p
W. G. Counsil RSB/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV ha t J
E Log # TXX-6414 P.
File # 10110 C
903.10
.r
=
Ref: 10CFR50.55(e) nlELECTRIC i
um %
April 30, 1987 Crecunw Vue Prrudem U. S. Nuclear Regulatory Commission ATTH: Document Control Desk Washington, DC 20555 4
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 SEISMIC DESIGN OF CONDUIT SDAR: CP-86-53 (INTERIM REPORT)
Gentlemen:
On October 21, 1986, we notified you by our letter, logged TXX-6054, of a deficiency involving the seismic design of conduit, which we deemed reportable under the provisions of 10CFR50.55(e).
This is an interim report submitted to
)
status corrective action implemented to date.
y The testing program to evaluate the ability of adhesives or clamps applied to
.i conduit couplings after installation to resist ~ torsional moments has been completed. The results indicate that the adhesives or clamps will not resist these torsional moments.
However, additional testing has begun on a heretofore unused combination of fiberglass cloth and resin which appears to be suitable for j
i the loading conditions identified based upon preliminary testing.
Formal testing I
of the fiberglass cloth / resin material is currently scheduled for completion by
' June 30, 1987.
The identification of further-corrective actions, including the
)
issue of revised design documents, is dependent upon the results of this testing.
1 We will submit our next report on this issue no later than September 21, 1987.
Very truly yours, W. G. Counsil By:
G. S. Keeley Manager, Nucle O u ing RSB/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 400 North Olwe Street L.B.81 Dallas Texas 75201
P Log # TXX-6450
- l :ll:
File # 10110 L
903.8
=
=
Ref # 10CFR50.55(e) 1UELECTRIC I$NNS$s,,,
May 15, 1987' U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 SUPPORT INSTALLATIONS - SPACING VIOLATIONS SDAR:
CP-86-59 (FINAL REPORT)
Gentlemen:
On August 29, 1986, we verbally Notified your. Mr. T. Westerman of a deficiency involving support installations with concrete attachments in violation of applicable CPSES design specifications.
Our latest interim report logged TXX-6109 was issued November 24, 1986.
This issue is applicable to Units l'and 2 and affects pipe, conduit, cable tray, and other support installations.
We have concluded the deficiencies identified above are reportable under the provisions of.100FR50.55(e) and the required information follows.
DESCRIPTION Engineering Evaluation Separation Violation (EESV) forms are used to document support spacing design problems.which require engineering resolution.
Modification to a support design resulting from the engineering resolution requires the initiation of the appropriate design change documents to implement the change.
Our evaluation has identified the following deviations from the established design change control program.
1.
EESV fonns have been issued for support installations without the required design change documents.
These installations were accepted by Quality Control (QC) based on technical dsta provided by the EESV form.
2.
Engineering disciplines have, in some cases, not controlled the issue / distribution of EESVs such that documented interdiscipline evaluation of all areas of the design and installation has not been assured.
3.
Procedural control of the EESV criteria has been inadequate.
400 North Olne Street I..B. 81 Dallas. Texas 75201
c s
-TXX-6450 May.15, 1987 Page 2 of 2 SAFETY IMPLICATION The improper use of EESVs has resulted in indeterminate design conditions whereby the function of the basic components (supports) can not be assured during seismic events.
. CORRECTIVE ACTIONS Procedure ECE-3.06-15 " Evaluation and Documentation of Concrete Attachment
. Spacing Violations" has been established to effectively control, transmit and document spacing violations. This procedure is scheduled for issue by June 15, 1987, and specifically assigns and clarifies engineering and inspection responsibilities.
To validate the support and structure' interface and to assure the integrity of the installations, an interdiscipline review will be i
performed.
In addition, QC will perform a 100% reinspection of the installations using revised design and design change documents.
The status of-any modification required as a result of this deficiency will be reported as corrective actions for the deficiency documented per SDAR: CP-86-04.
Very truly yours,
/
W. G. Counsil WLS/dl c - Mr. R. D. Martin - Region IV Mr. D. L. Kelley, RI - Region IV i
Mr. H. S. Phillips, RI - Region IV i
l
s, P_
Log # TXX-6434 4
l--
File # 10110
=
903.8
.__=
=
Ref. # 10CFR50.55(e)
TUELECTRIC wini.oi c. c oon.n Ltetutsve Viee fyrsident May 6* 1987 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION.(CPSES)-
DOCKET NOS. 50-445 AND 50-446 POLAR CRANE GIRDER AND GIRDER SUPPORT SHIM GAPS.
SDAR:.CP-86-60 (INTERIM REPORT) i Gentlemen:
On September 4,1986 we verbally notified your Mr. I. Barnes of a deficiency involving gaps between the polar crane girder support brackets.
This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e).
We have submitted previous interim reports logged TXX-6181.and TXX-6283, dated December. 18,.1986, and February 13, 1987, respectively.
')
Calculation. reviews which involve the resolution of Third Party comments are still in process. As a result, we are not able to complete our assessment of this issue at this time.
Our next report on this issue will be submitted no later than June 17, 1987.
Very truly yours, W. G. Counsil By:
G. S. Keeley Manager, Nuclear censing RWH/gj c - Mr. E. H. Johnson - Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 400 North Olive Street 1..B. 81 Dallas, Texas 75201
{
t J
7..
. w 8
a
. N-Log # TXX-6461
==
File # 10110 d
909.5 Ref # 10CFR50.55(e) 1UELECTRIC w n-c. non,w unwne ne nnaam May 22, 1987 i
U. S. Nuclear Regulatory Commission Attn:
Document Control Desk
'j Washington, D.C.
20555
)
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 INTERNAL C0ATINGS FOR UNIT 2 DIESEL FUEL OIL TANKS SDAR:
CP-86-64 (FINAL REPORT)
Gentlemen:
On September 4,1986, we verbally notified your Mr. I. Barnes of a deficiency involving the internal protective coatings for the Unit 2 diesel fuel oil storage tanks.
We have submitted interim reports, logged TXX-6013, TXX-6263, and TXX-6336, dated October 3, 1986, February 6,-1987, and March 23, 1987, respectively.
Our evaluation is completed and our conclusion is, had the l
issue remained uncorrected it would not have affected the safe operation of j
the plant. This issue is not reportable under the provisions of 10CFR50.55(e).
j Our investigation has determined that the original intent in invoking the requirements of Specification AS-31, " Protective Coatings," for the diesel fuel oil storage tanks was to ensure a high quality of workmanship through the specification's inspection and documentation requirements, not to classify the coating. However, once governed-by.the AS-31 specification for Safety Related containment coating, the tank coating became' regarded (incorrectly) as Safety Related.
In 1985, when this specification was declassified and containment coating became Non-Safety, the impact of this change on the tank coating was not addressed. These events led to the "... unresolved issues involving the repair process for the internal coatings of the Safety Related tanks" referred to in our letter TXX-6013.
We have concluded that the coating inside the fuel oil storage tanks is not Safety Related.
Our assessment considered coating degradation and potential blockage of the suction lines to the fuel oil transfer pumps.
Fluid flow rates in the immediate vicinity of the suction lines are too low (on the order of five feet per minute) to transport the large paint chips which would'be required to block the end of the suction line.
Should blockage of one suction line occur however, the redundant transfer pump and its suction line may be used to continue fuel transfer while the initial line blockage is cleared.
In addition, the day tank holds three hours of fuel below the low level alarm setpoint. Technical Specifications require the contents of the storage tanks to be sampled every 92 days for moisture and sediment.
It, the event that paint chips began to accumulate in the tank, the condition would be identified through these inspections and appropriate maintenance actions, as required by
' the Technical Specifications for CPSES, would be initiated.
400 North Olne Street I H Ni Dallas, lexas 75201
_ _ - - _ _ - _ _ - - _ _ _ ~
[ -( s..., y y
'TXX-6461-o May.22, 1987 Page 2 of 2 We.have additionally evaluated the cumulative effects of corrosion to the diesel generator fuel oil storage tanks.. Our evaluation determined that corrosion had been' considered in the. design of the fuel oil storage tanks.
The fuel oil tanks provide an additional wall thickness in excess of the one sixteenth' inch. (1/16"). corrosion allowance for the 40 year design life..
We have completed our assessment of Nuclear Regulatory Commission Information i
Notice-87-04 impact.on this issue and find it does not affect our evaluation.
We therefore conclude that neither the presence of failed coating particles or resulting~ corrosion could have adversely affected the safety of plant operation and the deficiency is not reportable per.10CFR50.55(e).
Records supporting this evaluation are available for your inspectors' review at the CPSES site.
Very truly yours, 4466f' W. G. Counsil RWH/gj-c - Mr. R. D. Martin - Region IV CPSES Resident Inspector - 3 copies 1
i
N 4
F====
Log # TXX-6445 FE File-# 10110 J
908.3 i
r
=
Ref # 10CFR50.55(e) 1 1UELECTRIC i
- "'"'C"""'
Emwve ike i+essdem May 15, 1987 U. S. Nuclear Regulatory Commission
~ Attn: Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 WEATHER PROTECTION FOR CLASS 1E COMPONENTS SDAR: CP-86-68 (INTERIM REPORT)
Gentlemen:
On October 2,1986, we verbally notified your Mr. I. Barnes of a deficiency regarding provisions for weather protection of Class IE components used in outdoor installations. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e).
Our most recent interim report was logged TXX-6226, dated January'21,1987.
Our evaluation involving weather proofing of outdoor Class 1E junction / terminal boxes is continuing. This program includes:
1.
Evaluation of engineering-field inspections conducted in accordance with CPE-FVM-EE-054, "FVM to Acquire Data on Weatherproofing of all Outdoor Class IE Terminal / Junction Boxes and Associated Conduit Fittings," and a review of previous inspection documents to assess the adequacy of previous inspections, 2.
The initiation of NCRs and the analysis of discrepant installations, 3.
Review of Specification 2323-ES-100, " Electrical Erection Specification", to assure adequate requirements for water-tight conduit connections for Class IE outdoor pull / junction boxes, and 4.
Review of applicable quality control procedures.
400 North Olive Street LB. 81 Dallas. Texas 75201
Me; j.
c.
.i '
.May 15s-l-1987
- Page 2 of 2' We will'_'be unable to' complet'e:our assessment ~ of this issueuntil the above
. activities are completed.l We anticipate submitting our next report by July. 24, 1987.
Very'truly yours, W. G. Counsil-WJH/gi-c -~ Mr. R.LD. Martin -- Region -IV Mr. D. L. Kelley, RI - Region IV Mr. H.'S.' Phillips', RI
. Region.IV h
l
-l l
.s
- 7 g
a E
Log # TXX-6464 L""" 9.
File # 10110 2
909.1
~~
C Ref
1UELECTRIC WiHimm G. Coun41 o,mov, n, nn,se,,,
May 29, 1987 U. S. Nuclear Regulatory Commission Attn: Document Control Desk s
Washington, D.C.
20555 1
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION-(CPSES)
DOCKET N05, 50-445 AND 50-446 SUPPORT OF CLASS 1E WIRING SDAR:
CP-86-69 (FINAL REPORT)
Gentlemen:
On October 2,1986, we verbally notified your Mr. I. Barnes of a deficiency
~
involving the failure to incorporate specific restraining / support requirements defined by IEEE Standard 420-1973, " Class IE Control Switchboards", into project specifications and procedures.
Our latest report, logged TXX-6406, was submitted on April 27, 1987.
Existing cable supports internal to Class IE Control Switchboards have been evaluated and found to be adequate. Also, an evaluation of similar cabinets has shown that this issue is not generic in nature. Wires internal to Class 1E Control Switchboards have been evaluated and found to be installed in accordance with the general practice guideline for good workmanship provided by IEEE Standard 420-1973 with the exception of seven cabinets. These seven cabinets had a portion of the wiring. supported vertically at 26 inch intervals instead of the 24 inch interval stated as the general guideline. A calculation (16345-EE(s)-131) has been generated concluding that unsupported spans up to 26 inches vertically are capable of resisting the maximum seismically induced tensile loads for this wiring. Terminations and butt splices were tested per UL 486C test standards to determine the minimum pull
)
i out forces.
Actual pull out forces for butt splices found during recent tests are far in excess of these minimums (4 to 5 times), and there is a direct relationship between termination and butt splice barrel design and installation methodology.
Calculations performed during the butt splice testing show the forces from seismic events to be less than both the minimum I
and tested pull out forces. The,refore, the adequacy of the termination and butt splices as a result of increased wiring support interval is not in question, so circuit continuity and function are not in jeopardy and this 1
issue does not represent a significant deficiency in design or construction.
j 400 North Ohve Street 1.. B. M t Dallas, Texas 75201
e
'TXX-6464-May 29, 1987 Page 2 of 2 A' Design Change Authorization (DCA-29,965) has been issued to revise Specification 2323-ES-100, " Electrical Erection Specification", to incorporate requirements equal to or greater than those specified in IEEE-420-1973 for-Class 1E wiring supports.
Corrective action will consist of rework, where
. j required, of _ Class IE Control ' Switchboards in accordance with-the DCA, after which required inspections will be performed by Quality Control.
We,have concluded in-the event this issue had remained undetected, no condition adverse to safety would have existed. This issue is not reportable under the provisions of 10CFR50.55(e).. Records-supporting this conclusion are available for your inspectors to review at the CPSES Project Site.
Very truly yours, y c l7tJC W. G. Counsil DAR/dl c - Mr. R.-D. Martin - Region IV CPSES Resident Inspector - 3 copies V
I
]
[s I
J M
Log > #.TXX-6372 d
==
File#:10110 i
d 908.3
=
=
Ref # IOCFR50.55(e) 1UELECTRIC April-27, 1987
. wmiam c. counsa Encune Vwe Prcudent
.U. S. Nuclear' Regulatory Commission 3
Attn:
Document Control Desk.
1 B
' Washington, D.C..20555
{
SUBJECT:
. COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
' DOCKET NOS. 50-445 AND 50-446 i
O
'ROCKWELL TERMINAL B0XES' SDAR: CP-86-74 (FINAL-REPORT)
Gentlemen:
On' October 21,-1986, we verbally notified your Mr. Ian Barnes of a deficiency j
regarding internal wiring discrepancies and rusty fittings in terminal boxes for main steam isolation valves supplied by Rockwell.
Our last interim report,_ logged TXX-6176, was submitted December 19, 1986. ' Verbal' extensions for submission of this report were made with your Mr. Lee Ellershaw on April' 10,-1987, and Mr. Cliff Hale on April. 20, 1987.
Specifically, the affected terminal boxes contain wiring for' limit switches for a' solenoid actuated valve and a manually ope' rated bypass valve associated with the Main Steam Isolation Valve (MSIV).
The manually operated bypass
' valve. utilizes the limit switches for position indication only'.
The manually operated bypass valve is only opened'to equalize pressure around the MSIV-prior to opening the first MSIV on plant startup.
Only one MSIV bypass valve will be opened at a, time.
The manual operation of these. valves during Modes 2, 3, and 4 will not significantly increase the ' potential for radioactive releases for postulated accidents requiring containment isolation.
This is
' based on:
1.
Lack of credible accident scenarios requiring containment isolation where MSIV bypass valves could significantly affect releases.
i 2.
Downstream valves (i.e., Turbine Stop valves, Auxiliary Feed Pump Turbine Steam Stop valves, Steam Dump valves and Steam Drain System) are closed, or are capable of remote manual closure in any containment isolation event.
3.
MSIV bypass valves are open for only a short time during each operating cycle. They will be locked closed during Mode 1 (power operation).
Therefore, the loss of position indication for the manual MSIV Bypass Valves will not have an adverse effect on plant safety.
Also, if necessary, the valve position can be locally verified.
400 North Olne Street LB RI Dallas, Tetas 75201
- 1
=wN:
,TXX-6372.
LApril 27,/1987f Page.2'of 2 The solenoid valve.liinit iswitches are part of the-MSIV. test. circuit and their
~
~failureLwill not prevent proper operation of the valve whether alshort or open 7
circuit-is postulated for the limit switches.
,J
- The conclusion of our evaluation-is' that, had the ' deficiencies remained.
1 uncorre'cted,..there would.have been no-adverse affect on the. safety of plant operations.
This, issue.is not reportable under the provisions of.
7
- 10CFR50'.55(e).
Nonconformance reports have been issued to document ands correct the discrepant. conditions.
Records supporting..this evaluation are available for your' inspectors' review at the CPSES site.
' ~
Very truly"yours,,,
/f/,i
'[lf]Os W. G. Counsil i
WJH/mlh~
.c - Mr. E. H.) Johnson - Region IV Mr. D. L'..Kelley, RI - Region IV 1,
Mr. H.-S. Phillips,'RI - Region IV j
4
s E
Log
'# TXX-6407 P9 File # 10110 i
.~"'
908.3
)
=
=
Ref #10CFR50.55(e) 1UELECTRIC j
EOU$T,s,,,,
April 29, 1987-U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 3
DEVIATIONS IN G0ULD BATTERY RACK DIMENSIONS SDAR: CP-86-76 (INTERIM REPORT)
Gentlemen:
On October 21, 1986, we verbally notified your Mr. I. Barnes of a deficiency i
involving the configuration and dimensions of Gould Battery Racks deviating j
from drawing requirements. This.is an interim report of a potentially t
reportable item under the provisions of 10CFR50.55(e). We have submitted an-interim report logged TXX-6110, dated November 20, 1986.
On April 22, 1987, we also verbally requested and received from the NRC an extension until May 1, 1987 to respond to this issue.
Class IE battery racks for Trains A and B, which were designed and fabricated by Gould, were found to have configurations and dimensions which deviated from the vendor design drawings.
These deviations could result in violation of battery cell spacing requirements for Class IE batteries.
Our review has determined that the conditions are applicable to Units 1 and 2.
Currently our evaluation is in the final stage of review.
Our next report will be submitted by June 5, 1987.
Very truly yours, M
W. G. Counsil RWH/gj c - Mr. E. H. Johnson - Region IV Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV 400 North Olive Street LB. 81 Dallas, Texas 75201
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Log # TXX-6487 P9 File # 10110
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2 903.9 i
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Ref # 10CFR50.55(e)
TilELECTRIC l
Wilham G. ( oumil i.m w = ne % wem June 4,'1987 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKETS NOS. 50-445 AND 50-446 STRUCTURAL STEEL BOLTING CONNECTIONS WITH THREADS IN THE SHEAR PLANE IN THE UNIT 1 CABLE SPREADING ROOM SDAR:
CP-86-78 (INTERIM REPORT)
Gentlemen:
On November 7,1986, we verbally notified your Mr. Tom Westerman of deviations in the structural steel bolting in the cable spreading. room.
This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e).
Our latest interim report, logged TXX-6326, was submitted on March 9, 1987.
Specifically, specification 2323-SS-168, Section 9.5, states that all field connections shall be made with high strength steel bolts in bearing type I
j connections with threads excluded from the shear planes except where otherwise l
l, stated or shown on engineering drawings.
Instances of bolted connections with L
threads in the shear plane which are not stated or shown on engineering l
drawings have been discovered in the Unit 1 Cable Spread Room.
)'
As previously advised (TXX-6326) calculations have been completed which verify design adequacy of Unit 1 Cable Spread Room Framing.
These calculations were prepared using the AISC Manual, 8th edition allowable bolt stresses and were found adequate.
However, the FSAR commits to the use of AISC 7th edition in which bolt stress allowables are more conservative.
An engineering review is in process to determine acceptability of the installation utilizing AISC'7th edition bolt stress allowables.
A review of the Design Change Authorization (DCA 32125, Rev. 0) allowing threads in the shear plane of Unit 2 Cable Spread Room Framing has been completed using the AISC 7th edition which found the allowable bolt stresses acceptable.
A review of 36 randomly selected sets of Units 1 and 2 Category I structural steel calculations revealeo 3 sets to be inadequate when threads of high strength bolts were assumed to be in the shear plane (AISC 7th edition stress allowables). As a result, all Category I and II structural steel calculations involving his trength bolts / connections are being reviewed assuming bolts have threads in the shear plane.
400 North Ohve Street I..B. 81 [)allas, Texas 75201
o
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A-1
'I TXX-6487
. June 4, 1987 l
Page 2 of 2-A review of calculations with bolt threads.in the shear plane for equipment foundations. has not -resulted in the identification of an ' adverse condition.
Under the. auspices of the equipment qualification program,.the original calculations are being checked for Safe Shut;down Earthquake (SSE) using Operational Bases Earthquake (0BE) design stress allowables, which are conservative.
If attachments are determined to be inadequate,' the bolts shall be replaced to eliminate threads from the shear plane using appropriate quality procedures and instructions for inspection and verification activities.
Evaluation of the effect of this issue on Cable Tray, Conduct, and HVAC Duct Supports is continuing and the-design adequacy of these systems will be included.in our next report.
Pipe Supports were evaluated and found to be acceptable.
They are designed in accordance with procedure CPPP-7 and are not governed by specification 2323-SS-168, directly.
'The determination of.this issue relative to the affect on the safety of plant operations during its 40 year design life is scheduled to be completed.by November 15, 1987.
Our next report will be' submitted by December 15, 1987.
Very truly yours,
.'). ;, b u rs -
W. G. Counsil 1
' 2 By:
.'J. S. Marshall Supervisor, Generic Licensing
'DAR/dl c - R. D. Martin - Region IV CPSES Resident Inspector - 3 copies m
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===r1===L Log # TXX-6480
- """ 9 File # 10110 908.3 2:
Ref # 10CFR50.55(e)
TUELECTRIC Hliham G. ( oumil
- i. mun, u a,mu,,,
June 5, 1987 U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 SEPARATION BARRIER MATERIAL ON POWER CABLES AND POWER RACEWAYS SDAR:
CP-86-83 (INTERIM REPORT)
Gentlemen:
9 On December 8,1986, we verbally notified your Mr. Ian Barnes of a deficiency involving the use of Separation Barrier Material (SBM) on power cables and power raceways.
Our last interim report submitted was logged TXX-6338, dated March 20, 1987.
We have concluded that this issue is reportable under the provisions of 10CFR50.55(e).and the required information follows.
DESCRIPTION Contrary to the established design requirements, thermal SBM was installed on a
power cables and associated raceways without considering the effects of the material on the ampacity of the cables.
As previously noted (in TXX-6338), calculations ha' e concluded that, due to v
significant conservatism in cable sizing design, the specific cables noted in the initiating nonconformance and cables in raceways did not exceed their insulation temperature rating during start-up testing.
Further evaluation of the deficiency, conducted to assure the cable sizing design is sufficiently conservative to counteract the derating effects of SBM, has resulted in the identification of 32 inadequate power cable installations.
These cables, listed in Attachment A, could have exceeded the ampacity rating previously assumed for plant operations (i.e., 40 year design life).
These cables are currently documented by Nonconformance Reports (NCRs).
4 SAFETY IMPLICATION SBN installed on power cables and/or power raceways requires derating of the cables' power carrying capacity (ampacity) to compensate for change in the operational environment.
By failing to address the derating factor, the power cables could exceed the design temperature rating.
Cable operation above the temperature rating is vipredictable. Actual failure of the cable insulation would be determined by 1.he degree of temperature increase, the time spent operating over the ratinc, and other factors - which makes the point of actual failure indeterminate.
However, failure could cause Class 1E power systems to become inoperative.
400 North Olhe Mreet I..B 81 Dallas. has 75201
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TXX-6480 June 5, 1987 Page 2 of 2.
This issue represents a significant final design deficiency as approved and released for construction and, were it to have remained uncorrected, could have adversely affected the safety of plant operations.
CORRECTIVE Ar.TIONS As a result of the analysis, NCR CE-87-4577 has been dispositioned to require removal of all'SBM from power cables / raceways except where specifically designated for radiant energy. shield purposes in the containment buildings.
To preclude additional instances of this deficiency, construction procedure CP-CPM 10.3, " Application of the Thermolag Fire Protection System", will be revised to adequately control the installation of SBM.
Our next report, statusing corrective actions, will be submitted by August 28, 1987. At that time, the construction schedule will be developed to the extent necessary to forecast actual completion dates.
Very truly yours,
'N W. G. Counsil WJH/d1'
. Attachment c - Mr. R. D. Martin - Region IV CPSES Resident Inspector - 3 copies l
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3
' Attachment-TXX-6480-l-
June'5, 1986 Page 1 of 1 Attachment A The following raceways contain the listed cables that would have insufficient
-ampacity for design life plant operations if B&B SBM were left in place:
Raceway No.
Cable No.
C12905099 A9100562 C12905984 A6100450 C12G05247 A$100562 C12G06343 EG100197A C02f12407 E9006954 C12513221 E$100818 C12918156
.E$100227A C12018157 E(100227B CO2918344 E9000376 C12G21190 EG105547 C22900418 E0200443, 444, 561 C22G05996 EG200194 T12$RBK06,07,08 A$100224Z T12KCBR14, 15 NK001103,'NK001133 T12KRBT23, 24, 25, 26, 27 NK101940Z-43Z, 46Z-48Z 512-53Z, 55Z-57Z.
T12KABN46 NK001178, NK101802, 803 i
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Log # TXX-6430 F9 File # 10110 903.11
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Ref # 10CFR50.55(s) n/ ELECTRIC Milliam G. Counsil w urowna mu+ni p,,
- 39g, U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555 i
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 ENVIRONMENTAL QUALIFICATION OF CABLE TO POST-LOCA HIGH RANGE RADIATION MONITOR SDAR: CP-87-05 (INTERIM REPORT)
Gentlemen:
l On April 6,1987, we verbally notified your Mr. C. Hale of a deficiency I
involving the failure of High Range Radiation Monitor (HRRM) coaxial cables to meet the minimum cable environmental qualification for dielectric insulation requirements. This is an interim report of a potentially reportable item under the provisions.of 10CFR50.55(e).
We have reviewed the Sorrento Electronics (SE) 10CFR21 Notification and have verified that it is applicable to CPSES eqelpment. This notification affects the following equipu nt for Units 1 and 2:
- 1) High range radiation monitors (1-RE-6290A&B; 2-RE-6290A&B)
- 2) Cables (EG0285112, EG028512Z, EG028682Z,.and EG028683Z) between the HRRM detector and the HRRM field mounted microprocessor. (RM-80).
The specific deficiency relates to the dielectric. insulation resistance required of the subject cables to allow the HRRM to operate within the guidelines established by Regulatory Guide 1.97.
The cable dielectric insulation resistance failed to meet minimum requirements.
I I
400 Nonh Olne Street LB. Hi Dallas has 75201
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W TXX-6430-May 6, 1987-Page 2 of 2.
H Our evaluation of this issue, with regard to the safety of plant operations,
(;.
is contingent upon the recommendation of corrective actions by SE.
While SE-has committed to provide the necessary corrective actions, they have not yet been received.
We anticipate submitting our next report on this issue no later than July 21, 1987.
Very truly yours,
- y. __
W. G. Counsil
/
By:
e e
G. S. Keeley Manager, Nuclear censing DAR/gj 4
c - Mr. E. H. Johnson - Region IV-Mr. D. L. Kelley, RI - Region IV Mr. H. S. Phillips, RI - Region IV c
p p
Log # TXX-6432 File # 10110 903.10 r
=
Ref: 10CFR50.55(e)
TUELECTRIC wnu.m c.cooma -
Mar 6, 1987 Eterwove e n e Presidem V. S.. Nuclear Regulatory Commission ATTN:
Document Control Desk
~ Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 SEISMIC QUALIFICATION OF'AS-BUILT 480 VOLT
. POWER CENTERS SDAR:
CP-87-06 (INTERIM REPORT)
Gentlemen:
On April 6,1987, we verbally. notified your Mr. Cliff Hale of a potentially i
reportable item involving 480 volt, 2000 KVA dry type transformers.
These components were originally seismically qualified by the vendor (Westinghouse) with the primary feed high voltage cables routed through the bottom of the. transformer enclo::ure.
In the CPSES installed configuration, the feed cables were routed through the top of the enclosure, possibly invalidating the seismic qualification.
We have developed a schedule of task description completion dates required to complete the evaluation to determine the possible impact upon the safety of plant operations. The schedule indicates the final report will be complete August 15, 1987.
We will submit our next report on this issue no later than September 4,1987.
Very truly yours, W. G. Counsil By:
v G. S. Keeley 1
Manager, Nuclear Li sing WJH/mlh c - Mr. E. H. Johnson, Region IV Mr. D. L. Kelley, RI - Region IV
];
Mr. H. S. Phillips, RI - Region IV 400 North Olive Street L H 81 Dallas. leus 75201
pi M
Log # TXX-6436 k
'= M File # 10110 1
h i
917.1 r
=
Ref:
10CFR50.55(e) 1 RIELECTRIC u n-c. coomu '
1 unwave na erniaem May 6, 1987
)
U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, DC 20555 j
I
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
)
DOCKET NOS. 50-445 AND 50-446 ITEMS PROCURED TO IMPROPER CODE CLASS SDAR:
CP-87-07 (INTERIM REPORT) 3
)
Gentlemen:
l On April 6,1987, we. verbally notified your Mr. C. Hale of a deficiency involving spare parts for safety related components.
This is an interim report cf a potentially reportable item under the provisions of 10CFR50.55(e).
Specifically, these components were procured with a Q-code "C" instead of a
'Q-code "A" as required. The documentation requirements imposed in these.
instances may not have been sufficient to assure that the quality of the item purchased was equivalent to the quality of the item as provided under the l
original equipment specification, or as required to maintain equipment qualification.
l The Spare Parts Review Program has been implemented to identify and resolve individual problems resulting from a generic deficiency in the procurement engineering review process as identified in Corrective Action Request 86-002.
The safety implications of the individual deficiencies (documented by site NCRs) are considered in addition to the availability of required documentation, installation status, and hardware replacement cost.
The Spare Parts Review Program is currently scheduled for completion by December 31, 1987.
The results of this program are required to determine the impact of this issue upon the safety of plant operations.
i 400 North Olive Street LB. 81 Dallas, Texas 75201
R" *' "._.'Mayl6,l1987f TXX-64361
=
Page 2 :.
ft We wil1 submit our next report on this-issue no later than February. 19, 1988.
7 Very'truly yours,
>>l' f.
4/
. (/V'.
. (.py &
~
W.'G.
ounsil
+
By:
~
G. S. Keeley' Manager, Nuclear'Li sing DAR/mlhL
.c; Mr. E._H. Johnson,~ Region IV
~
Mr. D.'L. Kelley, RIL. Region IV-Mr._H. S. Phillips, RI - Region IV i
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Log # TXX-6456 File # 10110 C
C 907.0.
r Ref # 10CFR50.55(e) 1UELECTRIC May 20, 1987 t,nume s ue erausem U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 CALIBRATION ACCURACY OF PRESSURE STANDARD (DEAD WEIGHT TESTER)
SDAR: CP-87-08 (INTERIM REPORT)
Gentlemen:
On' April 20, 1987 we verbally notified your Mr. Cliff Hale of a deficiency regarding'the accuracy tolerance of calibration for a dead weight' tester used as a standard in calibrating pressure instrumentation.
This tester.has been used in the Operation test / maintenance program. This is an-interim report of a potentially reportable item under the provisions of 10CFR50.55(e).
This issue is applicable to both Units 1 and 2.
'The safety significance of this issue will be determined based on the results of the.following action plan.
O All pressure instrumentation used in safety-related applications will be reviewed to determine the impact of calibration inaccuracies.
O Programs.important to safety (e.g., cold hydrostatic tests, preoperational tests, etc.) will be reviewed. for the effect of instrumentation inaccuracies, and 0
A review of other calibration standards such as temperature and flow will be performed to determine if generic calibration problems exist.
400 North Olive Street t.11. 8 i Dallas. lexas 75201
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sm Je 3 a4 J6 J
l
.TXX-6456...
M-
-May:20, 1987-Page 2 of 2 1
- r 1The above review
- isischeduled fo'r completion'by June. 30, 1987.
We will submit-Lour next: report on this' issue no later than July'21', 1987..
Very truly yours,.
W. G.-Counsil 9
By:
G. S.-Keeley-Manager, Nuclear.Lif(nsing' u.
- BSD/gj
. c Mr. E. H.' Johnson, Region IV Mr. D; L'. Kelley,-RI ( Region IV
- Mr. H..S. Phillips, RI - Region IV 9
l
' l, i
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gy
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Log # TXX-6495 M
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Ref # 10CFR50.55(e)
L
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TUELECTRIC wn. c.cooma June 5, 1987 Executhe Vke President B
U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKETS NOS. 50-445 AND 50-446 TERMINAL STUDS IN PX-2 TEST BLOCKS SDAR: CP-87-09 (INTERIM REPORT)
Gentlemen:
On May 7,1987, we verbally notified your Mr. Cliff Hale of a deficiency involving terminal studs in PK-2 test blocks.
This is an interim report of a 1
potentially reportable item under the provisions of 10CFR50.55(e).
TV Electric was notified by General Electric Service Advice No.182.1 of the susceptibility of terminal studs in PX-2 test blocks to stress corrosion cracking. These blocks are manufactured by General Electric for the CPSES 6.9kV switchgear. The susceptibility is due to the manufacturer's failure to 1
follow required manufacturing processes for the test blocks.
This deficiency is currently being evaluated by TV Electric to determine the impact of this issue upon the safety of plant operations. We expect to submit our next report on this issue by August 7, 1987.
l 1
Very truly yours, 4
44&nd W. G. Counsil WJH/gj.
j c - R. D. Martin - Region IV CPSES Resident Inspector - 3 copies 4
m I
400 North Olive Street L B. 81 Dallas, Texas 75201