ML20238A624
| ML20238A624 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/14/1987 |
| From: | Ellis J Citizens Association for Sound Energy |
| To: | Bloch P, Jordan W, Mccollom K Atomic Safety and Licensing Board Panel |
| References | |
| CON-#387-4243 OL, NUDOCS 8708210029 | |
| Download: ML20238A624 (61) | |
Text
{{#Wiki_filter:e .- 1 C A S E == 2.14/946 946 H (CITIZENS ASSN. FOR SOUND ENERGY) l
'87 AUG 17 All :07 August 14, 1987 ,p j
b0Cra <>, KL L b y i, w ) i Administrative Judge Peter B. Bloch Dr. Kenneth A. McCollom U. S. Nuclear Regulatory- Commission 1107 West Knapp Street Atomic Safety 6 Licensing Board Stillwater, Oklahoma 74075 Washington, D.C. 20555 Dr. Walter H. Jordan 881 W. Outer Drive i Oak Ridge, Tennessee 37830 f
Dear Administrative Judges:
Subject:
In the Matter of Texas Utilities Electric Company, et al. ) Application for an Operating License Comanche Peak Steam Electric Station Units 1 and 2 l Docket Nos. 50-445 and 50-446 Ob l '- 3 Potential 10 CFR 30.55(e) Items To easist the Board in its desire to be kept up-tr-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 a l received since we provided the Board with such items on July 8, 1987. l It is our intention to periodically continue to provide these SDAR's unless ; the Board indicates otherwise. I 1 Respectfully submitted, CASE (Citizens Association for Scund Energy)
/Ar6 % Y {y ylrs.)JuanitaEllis President l
cc: Service List, with Attachments B708210029 870014 PDR ADOCK 05000445 '. G PDR $ l _b56 o - - - - - - - - }
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M Log # TXX-6373 J
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' File # 10110
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i 903.9 r Ref # 10CFR50.55(e) 1UELECTRIC l wimun c. counsu ' Eteeme Vwe 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 HILTI-KWIK CONCRETE ANCHOR BOLTS SDAR: CP-80-10 (SUPPLEMENTAL % POP,T) Gentlemen: The subject issue addresses deficiencies resulting from unauthorized field modification of Hilti-Kwik Concrete Anchor Bolts. Specifically, this issue described modifications to Hilti. bolts in which the length of the bolts was reduced by crafts cutting off the mandrel portion of the Hilti-Kwik bolt and grinding a new mandrel into the shortened shaft, or crafts cutting off the-bottom mandrel of a double-mandreled Super Hilti-Kwik bolt. This deficiency was deemed reportable pursuant to the provisions of 10CFR50.55(e) and our final report was submitted in TXX-3442, dated December 4, 1981. This supplemental report is submitted as a result of further investigation into this issue. i Corrective Action for the deficiency involved an extensive program of , Ultrasonic Testing (UT). The results of this UT program were reported in our l final report. Among those results, it was reported that of the 796 Hilti-Kwik bolts examined in electrical installations, all were found to be acceptable l based upon not finding unauthorized mandrel cutting. 'l This letter clarifies those reported results in light of what appears to be conflicting statements in supporting documents. Several of these documents stated that 796 bolts in electrical installations were tested by UT examination and 5 were found to be unacceptable. The corresponding Nonconformance Report (NCR-E-81-00001) stated that those 5 nonconforming Hilti-Kwik bolts were unacceptable because they were shorter than indicated by the bolt stamps. The QC Inspection Report that was used to close this NCR noted that the bolt stamps appeared to have been altered. These five bolts were apparently not identified in our final report because they were outside the scope of the deficiency identified in SDAR CP-80-10, which specifically concerned cut mandrels. The 5 nonconforming Hilti-Kwik bolts identified in NCR-E-81-00001 were corrected by either removing the bolts or abandoning them in-place by driving the bolt into the hole to a depth beneath the concrete surface and then grouting the hole using an approved procedure. 400 North Obve Street LB 81 Dallas, Texas 75201
TXX-6373 July 15, 1987 Page 2 of 2 The circumstances and actions taken for the 5 unacceptable bolts were addressed in our response to unresolved item 50-445/8626-U-05; 50-446/8622-U-04 in accordance with the request in your letter dated April 2,1987 (see TXX-6472 dated June 5, 1987). Our conclusions regarding SDAR: CP-80-10 concerning the issue of cutting and regrinding of mandrels on Hilti-Kwik concrete anchor bolts remain unchanged. Very truly yours, b'b' Mid h W. G. C unsil 1 / ,j' i By: G. S. Keeley '" -
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Manager,NuclearLihg#nsing JCH/mlh .. c - Mr. R. D. Martin, Region IV i Resident Inspectors, CPSES (3 copies) 4
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.J i ', if Log # TXX-6588 s g
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,f, File # 10110 903.11 . p
._ .- 5 Ref. # 10CFR50.55(e) t 1UELECTRIC ;
[ , July 28,'1987i l i William G. Counsil ' Etuutne %ce Preudent c : U. S. Nuclear Regulatory Commission ' ' 4 ATTN: Document Control Desk ' Washington, D.C. 20555
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SUBJECT:
COMANCHE PEAK STEAM ELECTRIC. STATION (CPSES) ,; i ! i DOCKET NOS. 50-445 AND 50-446 ' t<
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QUALIFIED EQUIPMENT OUTSIDE CONTAINMENT s M. (<d
'j SDAR: CP-84-12 (INTERIM REPORT) ,1
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Gentlemen: On June 4,1984, we verbally notified your Mr. D. Hunnicutt of a deficienNh involving the environmental qualification of equipment outside containment .for high energy line breaks (HELB). This is an interim report of a potentially . 3 l reportable item under provisions of 10CFR50.55(e). Our latest interim report,6 i logged TXX-6482, was submitted on May.28, 1987. 1 C \ The evaluation of this issue with regard to safety of plant operations is continuing. We will submit our next report no later than September 28, 1987. j i
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Very truly your.s, 1 / l 1 ,
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W. G. Counsil By: < . 6 G. S. Keeley c- 3- .; Manager, Nuclear licarsing , u j BSD/amb c - Mr. R. D. Martin, Region IV ! Resident Inspectors, CPSES (3) j l l 1
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l 400 North Ohve Street LB 81 Dallas, Tens 73201 0
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===- - . Log # TXX-6555 n F M, File # 10110 '
.- 903.9
,, = = Ref #.10CFR50.55(e)
TUELECTRIC uly 6, 1987 y; .wmiam c. couma
.yl( .. r><cusava voce presuian, _ f! , U..S. Nuclear Regulatory Commission [' . 4 J 'a ATTN:' Document Control Desk pp > Wanipgton,D.C. 20555 j
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
+ DOCKET NOS. 50-445 AND 50-446
' CABLE TRAY' HANGER DESIGN -t SDAR: CP-85-35 (INTERIM REPORT) f"-
Gentlemen: On October 21, 1986, we notified you by our letter, logged TXX-6048, of a deficiency' involving .the design and construction activities of the cable tray )j support program which we deemed reportable-under the provisions of 10fflGO. 55(e) . This is an interim report submitted to status corrective j actfon implemented to date. Our latest interim report, logged TXX-6248, was j submitted on January 30, 1987. q We areitontinuing our evaluation and anticipate submitting our next report by l August' 3, .1987. . (
't t Very truly yours, l
lJ,d.Cw,nc'l i W. G. Counsil 1 By / b db J. S. Marshall l Supervisor, Generic Licensing l DAR/mgt
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c - R. D. Martin, Region IV Resident inspectors, CPSES (3) 400 North Ohve Street LB 81 Dallas. Tens 73201 __-_____-__-___-__-__--__-_N
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= N=- Log # TXX-6561 !
F9 File # 10110 903.9 r = Ref # 10CFR50.55(e) 1UELECTRIC William G. Counail ' beculac Yke Presnient U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 CABLE TRAY HANGER DESIGN ' SDAR: CP-85-35 (INTERIM REPORT)
- Gentlemen:
On October 21, 1986, we notified yon by our letter, logged TXX-6048, of a ' deficiency involving the design and construction activities of the cable tray-suppport program 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 latest interim report, logged TXX-6555, was submitted on July 6, 1987. Design verification based on the as-built conditions of the cable tray hangers is still in process, however, it is nearing completion. Static and cyclic tests to demonstrate. capacity to resist postulated loads have been completed for tray segments and fittings provided by T. J. Cope (Cope). Evaluation of these-test results is currently in progress. A second series,of tests will be conducted to evaluate Burndy Husky trays and fittings. The initial tests of trays with deviated splice plates for Cope installations have been completed. The results of these tests indicate that at least one (1) additional splice configuration may require testing. Testing of Burndy l Husky trays with deviated splice plates has been scheduled to begin upon the j completion of the Cope tray testing. j 400 North Olive Street LB 81 Dallas, Texas 73201 ________o
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TXX-6561 July 7, 1987 Page 2 of 2 Completion of all testing activities and the evaluation of the'results.is currently scheduled.for October 31, 1987. We anticipate submitting our next.
. report.on this issita'no later than November 19, 1987. i Very truly yours, Ibn W. G. Counsil I By: .
/.'/
G..S. Keeley .W Manager, Nuclear. Lice # sing-DAR/mgt
-c'-:R. D. M'artin, Region IV Resident Inspectors,.CPSES (3) 1 I
l l I
l i l M Log # TXX-6556 -
-- C Fi1e # 10110 i
- j 903.9 j l
= = Ref # 10CFR50.55(e) l 1UELECTRIC.
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- wunam c. couma "I# 'I Enceutive %ce President 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 TRAIN "C"' CONDUIT SUPPORTS SDAR: CP-85-36 (INTERIM REPORT)
Gentlemen: On September 6, -1986, .we verbally lotified your Mr. C. Gould of a deficiency . involving the-installation of non-safety related conduit (two-inch diameter and smaller)Lon non-safety related dead weight supports in Category 1 areas. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Our latest interim report was logged TXX-6221, dated January 21, 1987. As previously reported, the evaluation program developed to address this issue
-(CPRY Program Plan ISAP I.c) is still in process.
Currently, our field verification is continuing in the Unit I and Common Area , to determine the adequacy of the Train C' conduit supports. To date, the evaluation of conduit supports in 180 rooms has been completed. These efforts H represent approximately 63-percent of the rooms and 72 percent of the supports in the Unit I and Common Area. Evalu;iion and verification activities for Unit.2 will be addressed after the completion of the Unit 1 and Common Area activities. t
.Our~ evaluation of the Unit-1 and Common Area is currently scheduled for completion by late December,1987. We anticipate submitting our next report on this issue no later than February 10, 1988.
Very truly yours, i [ W. G. Counsil By: '
/
G. S. Keeley er / . Manager,NuclearLicen/ing DAR/gj c - R. D. Martin, Region IV Resident Inspectors, CPSES (3) ! 400 Nonh Obve Street LB 81 Dallas, Texas 75201 '
M Log # TXX-6595 F9
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File # 10110 903.9
= = Ref # 10CFR50.55(e)
TUELECTRIC William G. Counsil EAccutne Vwe Pmukm N. S. Nuclear Regulatory Commission g
.At vi: ' Document Control Desk- '
Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET N05. 50-445 AND 50-446
. THERM 0 LAG ON CONDVIT SUPPORT
~SDAR: CP-85-42 (INTERIM REPORT)
Genticmen: On October 26, 1986, we notified you, by our letter logged TXX-6049, of a
. deficiency involving the possible adverse effect of the substitution of rectangular and over-sized preformed sections of thermolag on conduit '
installations which we deemed to be reportable under the provisions of 10CFR50.55(e). This is an interim report submitted to provide the status of our corrective actions. Our last interim report was logged TXX-6420, dated April 28, 1987. The implementation of corrective actions for this issue is continuing. Our next report on this issue will be submitted no later than September 11, 1987. Very truly yours, b & 1Ut W. G. ounsil
/ 4 /
By: e G. S. Keeley bL$f c- / Manager, Nuclear Liceysing DAR/dl c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3)
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M Log # TXX-6613 F "-- File # 10110
.-- 903.8 r = Ref. 10CFR50.55(e) 7UELECTRIC August 3, 1987 William G. Courull Encuene Vwe Prrskient 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 PROXIMITY PLACEMENT - RICHMOND INSERTS TO EMBEDDED PLATES AND WELDED ATTACHMENTS TO EMBEDDED PLATES SDAR: CP-86-04 (INTERIM REPORT) Gentlemen: On February 13, 1987, we notified you by our letter logged TXX-6260, of 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). Our last interim report was logged TXX-6423, dated May 1, 1987.' Our corrective actions previously specified at a continuing as planned. Since our last report, spacing criteria issues arising from the internal review of
' Design Specification 2323-SS-30 (Structural Embedments) have been resolved and quality control instructions have been added as an appendix to the specification. This effort has delayed the issuance of the specification revision which is expected to be issued shortly. Following the issuance of this specification, engineering will complete their evaluation to identify all corrective actions and means to prevent recurrence for this and related issues, as described in our letter logged TXX-6260, dated February 13, 1987.
We anticipate submitting our next report no later than October 5, 1987. Very truly yours, L 'Jt; W. G. Counsil BSD/gj c - Mr. R. D. Martin, Region IV ! Resident Inspectors, CPSES (3)
.M) Nonh Obve Street LB MI Dallas, Texas 7$201
, Log # TXX-6556 '
.9 ~ ' " '
File # 10110 l
- - 909.2 l r- C Ref # 10CFR50.55(e) i 1UELECTRIC wmiam c. coumn July 15,1987 Execuche %ce Presuknt 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 SERVICE WATER SYSTEM LEAKAGE SDAR: CP-86-07 (INTERIM REPORT) Gentlemen: On April 11, 1986, we notified you by our. letter logged TXX-4762 of a
~ deficiency involving weld failure and coating degradation in the Service Water.
System which we deemed reportabl.e under the provisions of 10CFR50.55(e). This is an interim report.. submitted to status corrective actions implemented on Unit 1 to date. Our latest interim report, logged TXX-6348 was submitted on March 26, 1987. i As previously reported, the field weld analysis has been '.ompleted and t currently the results are under engineering evaluation.
.In our last report we advised that SWEC Engineering (Boston) was performing a corrosion study which was scheduled to be completed by June 15, 1987.
Completion' of ~this study has been rescheduled for August 26, 1987. We anticipate submitting our next report on this issue by September 30, 1987. Very truly yours, ! f, W W. G. Counsil l By: b [ G. S. Keeley - ' > 1 Manager,NuclearLyensing DAR/gj c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 Nc.th Ohve Sucet LB 81 Dallas, Texas 7D01 _____________________b
M Log # TXX-6591 F "-- File # 10110 ) _ . 903.9 i'
.= = Ref # 10CFR50.55(e)-
illELECTRIC 4 wim m c. counsu July 22, 1987 Executu>e Vwe Pres &nt 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 INSTALLATION OF SUPER HILTI KWIK BOLTS SDAR: CP-86-08 (INTERIM REPORT) l Gentlemen: ' On February 5, 1987, we verbally notified your Mr. T. F. Westerman of a deficiency involving the control of stamps used to identify Super Hilti Kwik-Bolts (HKB)._ This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Our previous report was logged TXX-6318 dated March 12, 1987. L 1 Our evaluation for the possible adverse effects of these conditions on plant safety is continuing in accordance with the site initiated corrective action ; request CAR-074. - i Our next report on this issue will be submitted no later than October 22, 1987. 2 Very truly yours, SY W. G. Counsil By: A.- "
"G. S. Keeley C'-
Manager,NuclearLkensing i GLB/gj I c - Mr. R. D. Martin, Region IV
. Resident Inspectors, CPSES (3) 1 l
400 North Olive Street LB 81 Dallas, Texas 73201
l. L i
-- Log # TXX-6612 ' I FE File # 10110 908.1' 1
= r Ref. # 10CFR50.55(e) 1UELECTRIC wmim c. coumn July 30, 1987 nuuma v,a naacm ;'
.U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 i
l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS, 50 445 AND 50-446 l ' ELECTRICAL PENETRATION ASSEMBLIES SDAR: CP-86-10 (INTERIM REPORT) Gentlemen: l
'On April-24, 1986, we notified you by our letter ic,gged 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). Our last interim report, logged TXX-6467, was submitted on June 1,1987. The schedule' for corrective actions relating to Enforcement Action (EA) 86-09 EPA module replacement has been revised to September 30, 1987 for Unit 1 and flovember 30, 1987 for Unit 2. This revised schedule for completion is a result of recently established approval requirements for any welding being performed to the containment liner. Associated components of the Medium Volt Power (MVP) assemblies and Nuclear Instrumentation Systems (NIS) modules require welding to the containment liner to facilitate installation. ! We expect to submit our next report by December 18, 1987. Very truly yours, hV, -tcwf W. G. Counsil By: ( ; G. S. Keeley ' Manager, Nuclear Li g
.y BSD/dl c - Mr. R. D. Martin - Region IV Resident Inspectors, CPSES (3) 300 North Olive Street LB 81 Dallas, Texas 73201
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[ l i Log # TXX-6606 File'# 10110 , 910.3 l M Ref # 10CFR50.55(e) L Y%
, TUELECTRIC 0*11 0' 198 I
William G. Counsil Etemene he Premwm 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 INSTRUMENTATION INSTALLATIONS ! SDAR: CP-86-19 (INTERIM REPORT) Gentlemen: ! On March 21, 1986, we notified you of a reportable item involving the installation of steam-service pressure transmitters (see TXX-4824 dated May 23, 1986). This is a follow-up interim report to status the corrective actions for a reportable item under the provisions of 10CFR50.55(e). Our last interim report was logged TXX-6332 dated March 20, 1987. The scope of this item has been expanded to include corrective action implementation for SDARS: CP-86-16, " Fire Effects on Instrument Tubing," CP- ' 86-50, "Unistrut Spring Nuts on Instrument Supports," CP-86-70, " Elevated Temperature Effects on Instrument Supports and Tubing," and CP-86-77,
" Instrument Tubing Minimum Wall Thickness"; all of which were previously reported under the provisions of 10CFR50.55(e).
.SDAR CP-86-16: FIRE EFFECTS ON INSTRUMENT TVBING Engineering evaluation of this issue is continuing; corrective actions have yet to be defined.
SDAR CP-86-19: INSTRUMENTS MOUNTED ABOVE PROCESS TAPS Engineering is developing the design modification package (DM-86-140) required to modify the Unit 1 instrument installations. This design modification will involve relocating the instruments to an elevation below the root valves and rerouting the tubing to the proper slope requirements as defined in Specification CPES-I-1018. Unit 2 instruments will be corrected similarly to Unit 1. l ( 400 North Obve Street LB 81 Dallas, Texas 75201
g . TXX-6606 July. 30,1987 Page 2 of 3 ; SDAR CP-86-50: UNISTRUT SPRING NUTS ON INSTRUMENT SUPPORTS The following is the corrective action plan for this issue. 1
- 1) Construction shall revise the installation procedures to reflect the new criteria (torque and alignment) as defined in Specification CPES-I-1018.
i
- 2) Construction personnel will be trained in the new criteria, i 1
- 3) Quality Control shall revise the inspection procedures. to reflect i the new criteria as defined in Specification CPES-I-1018. 1
- 4) Quality Control personnel will be trained to the revised I procedures.
- 5) Construction personnel shall rework 100% of the installations to l conform to the new criteria with Quality Control inspection ^
personnel providing verification and documentation of torque and ; spring nut alignment, j. SDAR CP-86-70: ELEVATED TEMPERATURE EFFECTS ON TUBING AND SUPPORTS Engineering has reviewed and evaluated calculation number 304, Rev. 1, " Review of Accident Elevated Temperature Effects on Tubing," and concludes that any 4 additional loads due to accident conditions are included in the tubing support criteria. The calculation incorporates strain criteria as suggested in NUREG 1061, " Report of the US NRC Piping Review Committee," Volume 2, Section 2.5. Safety-related instruments and non-safety related instruments connected to ASME III fluid systems and ANSI safety class installations associated with ' CPSES Unit 1 and 2 as defined by field verification method CPE-FVM-IC-069 are being evaluated in accordance with this criteria. SDAR CP-86-77: INSTRUMENT TUBING MINIMUM WALL THICKNESS The following is our corrective action plan for this issue.
- 1) The minimum wall thickness of 0.035 inches has been determined to be acceptable for 3/8 inch 0.D. stainless steel tubing in all i applications.
- 2) Quality Control shall revise the inspection procedures to reflect the new criteria provided in Specification CPES-I-1018 for all 1/2 i inch 0.D. stainless steel tubing installed as impulse lines in the reactor coolant, chemical and volume control, and safety injection systems.
- 3) Inspection personnel will be trained in accordance with the revised procedure.
- 4) QC will perform 100% reinspection per revised procedure and document all deficiencies via NCRs.
, ;g..- TXX-6606'
' July 30, 1987 Page 3 of 3 l
-5) Engineering will evaluate each deficiency and disposition the NCR.
It-is anticipated that rework will be required of any deficiencies ;
' found in the reactor coolant system tubing. Chemical and volume '
control system and safety injection system deficiencies will be evaluated with due consideration for actual pressure / temperature exposure.
- 6) Construction shall revise the existing procedure to reflect the new criteria of Specification CPES-I-1018.
- 7) -Construction personnel will be trained in accordance with the revised procedure, l i
- 8) Construction will rework the deficiencies as directed by I Engineering to conform to Specification CPES-I-3018.
t
- 9) Rework will be reinspected for compliance to the revised' procedure. !
Engineering Calculation #16345/6-IC-(B)-001 was prepared to determine the maximum pressure / temperature conditions which 0.035 inch thick wall stainless steel tubing may be exposed. This calculation determined the minimum wall thickness required for the maximum pressure / temperature anticipated at CPSES. ' The criteria derived and based on this calculation has been incorporated into Specification CPES-I-1018. Engineering evaluations of SDAR CP 86-16 and CP-86-70 are ongoing. Resolution and corrective actions are scheduled to be completed by October 30, 1987. J Our next report will be submitted by December 2, 1987. i Very truly'yours, l W. G. Counsil !
,/ .1 By:
G. S. Keele s . L
/
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Manager,Nucleapicensing GLB/gj c - Mr. R. D.' Martin, Region IV 1 j Resident Inspectors, CPSES (3) l i
l l I n l LA "
-- Log # TXX-6551
---- E
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File # 10110 909.4
= = Ref # 10CFR50.55(e) illELECTRIC wmam c. counsu uly 8, 1987 i
E.necutne %ee Presubent l 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) Gentlemen: l ' On April 2,1986, we verbally notified Mr. T. F. Westerman of a deficiency involving the capacity of the present fan cooler desi to adequately handle the heat load in several rooms. gnOur in the Safeguards latest interim area report, TXX-6459, was submitted on May 20, 1987. We have concluded that this l deficiency is reportable for Unit I under the provisions of 10CFR50.55(e) and i the required information follows. Our evaluation of the deficiency for Unit 2 is continuing. Description of Problem 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 1 coolers was accepted at values below both the design beat loads and equipment capacities. An engineering evaluation, undertaken to ensure equipment cooling capacities meet the required design criteria, has concluded the emergency fan coolers l (served by the safety chilled water system) are adequately sized to maintain room temperatures, within the FSAR requirements. However, system modifications are required to rebalance the existing chilled water flow rates. Safety Implication In the event the condition had remained undetected and the original flow rate distribution uncorrected, the deficiency could have resulted in an inability of associated safety-related systems to perform as required iollowing a LOCA with loss of offsite power. I 400 Nonh Ohve Street LB BI Dallas, Texas 7 DOI
2 h w< TXX-6551 July 8, 1987 Page 2 of 2 Corrective Action The engineering evaluation to rebalance the flow rates for Unit I has been completed and the necessary revised design documents provided to Operations and Start-Up for implementation. Our schedule to accomplish this task will be formulated no later than September 30, 19'37. Currently, project procedures are under evaluation to determine corrective actions required to prevent recurrence of inadequate test acceptance criteria. This portion of our evaluation is scheduled to be completed by September 30, ; 1987. Unit 2 calculations are scheduled for completion by December 31, 1987. Upon completion of these calculations and the associated evaluations, our assessment of this issue for Unit 2 can be completed. We anticipate submitting our next report no later than October 21, 1987. Very truly yours,
, LYk b W. G. Counsil '
By: _. . G. S. Keeley /- i Manager, Nuclear Li'psing GLB/mgt c - Mr. R. D. Mi.rtin, Region IV Resident Inspectors, CPSES (3) l 4 l l l _ . . .___._______m_.________
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M
~~ Log # TXX-6564 !
File # 10110
-. .-- . 909,5 l
= = Ref # 10CFR50.55(e) 1UELECTRIC William G. Counsil '
Execme We President U. S. Nuclear Regulatory Commission ; ATTN: Document Control Desk Washington, D.C. 20555
SUBJECT:
. COMANCHE PEAK STEAM ELECTRIC STATI0fl (CPe:S)
DOCKET NOS. 50-445 AND 50-446 DIESEL GENERATOR LUBE OIL SUMP TANK FOOT VALVES I SDAR: CP-86-31 (FINAL. REPORT) l 1 Gentlemen: On August 22, 1986, we notified you by our letter, logged TXX-4975, of a-deficiency-involving the possible failure of the rubber facing on foot valves mounted in the diesel generator lube oil sump tanks, which we deemed reportable under the provisions of 10CFR50.55(e). This is a final report submitted to status corrective actions. Installation of the Unit One replacement rubber facings is complete. The Unit Two replacement rubber facing installation is currently scheduled for l completion by December 30, 1987. i , .Very truly yours, h , ) W. G. Counsil ' By: /
-G.
S. Keele
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(( I Manager, Nuc ea ~i ensing RWH/mgt . l c - R. D. Martin, Region IV Resident Inspectors, CPSES (3) ' i i i 400 North Olive Street LB 81 Dallas, Texas 7320) t
M Log # TXX-6622 FE File # 10110 903.9
= = Ref # 10CFR50.55(e) 1UELECTRIC winiam c. counsii August 4, 1987 Esecuuve Vwe Presnkat U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE.PEAX STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 MOTOR OPERATORS FOR MANUAL VALVES SDAR: CP.-86-35 (INTERIM REPORT) I 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 May 15, 1987. We.are continuing corrective action implementation and anticipate submitting our next report on this issue no later than October 2, 1987. 1 Very truly yours, ! Ld. b - W. G. Counsil By: D. R. Woodlan Supervisor, Docket Licensing JDS/mgt c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3)
- 40) Nonh Olive Street LB 81 Dallas. Texas 73:01
M i.og # TXX-6586 l3
- " "::: . File # 10110 i 908.3-n 33~ 133 4
Ref # 10CFR50.55(e). I inWELECTRIC i Willismi G. Coumil July 24, 1987 Enecurne %ce Premkm ! 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 i CLASS lE BATTERY CHARGER COMPONENTS SDAR: CP-86 37 (INTERIM REPORT) !
. Gentlemen: ~
, t 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-6543, dated June 26, 1987. We are continuing our evaluation and anticipate submitting our next report by September 4, 1987. Very truly yours, , ik W. G. Counsil VIP / ij i c - f. D. Martin, Region IV , Rasident Inspectors, CPSES (3) t i 400 Nonh Ohve Street LB 81 Dallas. Tem 73201 _ _ _ _ _ _ - _ ___.___--____-_--_Q
M Log # TXX-6560
" FE 1 -
Flie # 10110 10110.1 i r = Ref # 10CFR50.55(e) nlELECTRIC wmiam c. coumii July 15, 1987 bxutne he Presusent i 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 ADEQUACY OF NON-CONFORMANCE DISPOSITIONS SDAR: CP-86-48 (INTERIM-REPORT) Gentlemen: On June 16, 1986, we verbally no,tified your Mr. C. Hale of a deficiency involving the adequacy of nonconformance dispositions. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(.e). Our latest interim report on this issue, logged TXX-6036, was submitted on-October 20, 1986. i Gur previous letter stated that we planned to perform a technical review of all previously closed Non-Conformance Reports (NCRs) with " void", " repair", or "use-as-is" dispositions and complete this review by June 30, 1987. In order to more e4ficiently utilize the resources available to the project, we have found it necessary to reschedule the completion of this program. The complete NCR review program consisting of the initial screening, technical evaluation, and resolution of safety related concerns, is now scheduled for completion by fuel load. The following is submitted to provide a more detailed description of the technical review program for NCRs. The NCR Review Program is defined per procedures CPE-TD SWEC-034, Rev. 1, "Non-Conformance Report (NCR) and Design Deficiency Report (TDDR) Evaluation," and PP-041, Rev.1, "Non-Conformance Evaluation Procedure." The scope of this review program encompasses only those documents issued before the corporate procedure NE0 3.05. " Reporting and Control of Non-Conformances"(effective date: December 22, 1986) was issued. This program is comprised of three ! phases: 400 North Olive Street LB 81 Dallas, Texas 73201
g.' ,. P L TXX-6560 l July 15, 1987 Page 2 of 2 ) Phase I i
. Identifies NCRs containing dispositions which,'. based upe, !
an engineering assessment of the actual written 1
-dispositions, affect engineering requirements (i.e., use- {
as-is, repair, or void) and are not the subject of other engineering validation. efforts. These validation efforts l include: l
- a. Pipe Support Engineering (SWEC) !
- b. Equipment Qualifict. tion (Impell/SWEC)
- c. Cable Tray Hangers (EBASC0/Impell)
- d. Conduit (EBASC0/Impell)
- e. HVAC (EBASC0/Impell)
Phase II Consists of a detailed technical evaluation of the NCRs identified in Phase I. Phase III consists of the resolution of disposition issues, as required, bas,ed on the Phase II Review. Each individual NCR addressed during Phase III of this program will be . reviewed for safety significance in accordance with 10CFR50.55(e). Our assessment of the deportability of this issue will result from this review. Our next report on this issue will be submitted by November 20, 1987. Very truly yours, W. G. ounsil By: < G. S. Keeley " GLB/mgt Manager,NuclearLYensing c - Mr. R. D. Martin, Region'IV Resident Inspectors, CPSES (3) 4
,k
==== - Log # TXX-6552
- M E File # 10110 903.9
= = Ref # 10CFR50.55(e)
TUELECTRIC wmiarn c. coun3n July 8, 1987 E ecutwe Vxe Preskknt U.' S. Nuclear Regulatory Commission ' ATTN: . Document Control Desk Washington, D.C. 20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) : DOCKET NOS. 50-445 AND 50-446 ANCHOR BOLTS SUPPLIED BY HILTI SDAR: CP-86-51 (INTERIM REPORT) Gentlemen: On July 17, 1986, we verbally notified your Mr. I. Barnes of a deficiency involving a condition cited at another. nuclear facility in which anchor bolts supplied by Hilti do not meet average ultimate tensile loads in certain sizes as published by the supplier. Our latest interim report, logged ,TXX-6405, was l submitted on April 24, 1987. l Our current evaluation of this issue has considered the revised allowable loads and those used in the design with associated safety factors. This comparison indicates 1/2" diameter bolts in specific are not acceptable when , l acknowledging published average ultimate tensile loads. In order to assure j l the adequacy of these bolts, testing of the 1/2" diameter anchor bolts will be ; performed in CPSES concrete to determine the ultimate capacity. I A procedure is being developed for the testing activity to be performed. This procedure is currently scheduled for completion by late September, 1987. Our next report on this issue will be submitted no later than October 21, l 1987. Very truly yours, W. G. C unsil , By: < ' L/ G. S. Keeley % / Manager, Nuclear L,1 censing GLB/mgt c - R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 North Ohve Street LB!! D?"as. leus 75201
. i n
i-M Log # TXX-6618 FE File # 10110 903.11 l
== :: Ref: 10CFR50.55(e) !
1UELECTRIC wmiam c. coumn August 3, 1987 Executsve Vwe hmJent J. S. Nuclear Regulatory Commission i Attn: Document Control Desk Washington, D.C. 20555 ,
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 J l On April 29, 1986, we notified you'by our' letter, logged TXX-6415, that the deficiency' involving splice / connections used for cable trays in Units 1 and 2 was reportable under the provisions of 10CFR50.55(e). Our evaluation of this issue is continuing. We expect to submit our next report by September 3,1987. Very truly yours, 3 l7Q& ' W. G. Counsil RSB/mlh < c - Mr R. D. Martin, Region IV Resident Inspectors, CPSES (3) A 400 North Olive Street LB 81 Dallas, Texas 73201 b ;_
'l M Log # TXX-6581 !
F9 File # 10110 903.7 r C Ref: 10CFR50.55(e) illELECTRIC wmi m c. counsa July 21, 1987 Executne Vwe Presukat U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET N05. 50-445 AND 50-446 ORIGINAL DESIGN OF CONTROL ROOM CEILING SDAR: CP-86-54 (INTERIM REPORT)
-Gentlemen:
i On October 3,1986, we notified you by our letter lagged TXX-6007 of a deficiency involving the, design and installation of the control room ceiling which we deemed reportable under the provisions of 10CFR50.55(e). This is an interim report submitted to status corrective actions implemented to date. Our most'recent interim report was logged TXX-6291, dated February 20, 1987. Resolution of the open Quality Control inspection items is continuing. We will submit our next report on this issue no later than September 21, 1987. Very truly yours, k, ML 'h W. G. Counsil By: / / G. S. Keeley / Manager, Ntfebr# Licensing JDS/mlh c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 North Olive Street LB 81 Dallas, Texas 75201
M Log # TXX-6553 FM .- File # 10110
. 903.7
= = Ref # 10CFR50.55(e)
TUELECTRIC wmiun c. counsii July 8, 1987 E.necutae Vwe Pmident U. S. Nuclear Regulatory Commission Attn: . Document Control Desk ! Washington, DC 20555 j
.)
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 SEISMIC AIR GAP DESIGN ADEQUACY SDAR: CP-86-55 (INTERIM REPORT) Gentlemen: On' August 14, 1986, we verbally notified your Mr. I. Barnes of a deficiency involving the. seismic air gap design ade and Units 1 and 2 Containment Buildings.quacy This is between an interim the Auxiliary report ofBuilding a potentially reportable item under the provisions of 10CFR50.55(e). Our last i interim report was logged TXX-6321, dated March 13, 1987. The evaluation of the calculations indicating acceptability of the existing j gap Buildingis continuing Analysis. in Alternate conjunction with the validation of the Seismic Category I 1 calculations are currently being developed. 1 After the completion of these alternate calculations, a conclusion will be I made regarding the safety of plant operations had this condition gone uncorrected. ; y We are continuing September 11, 1987.our evaluation and anticipate submitting our next~ report by i Very truly yours,
$b .
tAf W. G. Counsil By: / M / G. S. Keeley " ' JCH/mgt Manuger, Nucl:arIicensing ' l t c -Mr. R. D. Martin, Region IV i Resident Inspectors, CPSES (3) l 4 400 Nonh Olive Street LB 81 Dallas, Tem 73201
M Log # TXX-6573 FE File # 10110 903.9 r = Ref: 10CFR50.55(e) i 1UELECTRIC wmiam c. counsa July 20, 1987 Emutne Vwe Pressknt U. S. Nuclear Ragulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 PIPE SUPPORT INSTALLATIONS. SDAR: CP-86-63 (INTERIM REPORT) Gentlemen: By letter logged TXX-6027, dated October 16, 1986, we notified you of a reportable item involving pipe support construction deficiencies. This is a follow-up interim report on a reportable item under provisions of 10CFR50.55(e). Our last interim report was logged TXX-6391, dated April 16, 1987. Previous reports described the extent of the Hardware Validation Program (HVP) - for the physical inspections of large and small bore safety related piping for Unit 1. In conjunction with this program, an additional list of attributes has been formulated for physical ' inspection. These attributes, identified as the Supplemental Inspection Checklist (SIC), are the result of requirements for additional information due to changes in the engineering analysis method or items deemed prudent and/or conservative to reinspect in concert with the HVP. Procedure CP-QAP-12.1, titled " Mechanical Component Installation Verification," has been revised to implement the SIC requirement. Any l deficiencies discovered during final inspections shall be documented via an ' NCR and transmitted to Engineering for resolution and assignment of appropriate actions. Seven (7) percent of the approximately 18,640 supports involved in the Unit 1 and common HVP and SIC reinspection have been completed. Our estimated completion date is now January 1,1988. l l
\
400 North Olive Street LB 81 Dallas, Teus 75201 i _________--__-_A
i TXX 6573 July 20, 1987 Page 2 Our next report will be submitted on or before September 30, 1987. Very truly yours, Ni l. W. G. Counsil By: k k / G. S. Ke'eley c. Manager, Nuclear nTing BSD/mlh-c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) I l 1
I M i.og # TXX-6590
,= 9 File # 10110
.- 908.3 r = Ref # 10CFR50.55(e) 1UELECTRIC \
wmiarn c. counsit July 24, 1987 Esentove Voce Preudent ' U,.S. Nuclear Regulatory Commission Attn: Document Control Desk l Washington, D.C. 20555 i
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 1E components used in outdoor installations. This is an interim report of a potentially reportable i item under the_ provisions-of 10CFR50.55(e). Our most recent interim report was logged TXX-6445, dated May 15, 1987. -
We are continuing our evaluation and anticipate submitting our next report by September 4, 1987. j Very truly yours, [/'752 W. G. Counsil
-VIP /gj c - R. D. Martin, Region IV Resident Inspectors, CPSES (3)
~ .
400 Nonh Obve Street LB 88 Dallas. Teus 73201 1
M Log # TXX-6636 F -- File # 10110 3
.-- 908.3
= = Ref # 10CFR50.55(e)
TUELECTRIC william c. counsit Eucuttve %ce Vreudent August 7, 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. DEVIATIONS IN GOULD DATTERY RACK DIMENSIONS SDAR: CP-86-76 (INTERIM REPORT) Gentlemen: On October 21, 1986, we verbally notified your Mr. I. Barnes of. a deficiency < involving the configuration and dimensions of Gould Battery Racks deviating from drawing requirements. This is an interim report of a potentially 1-reportable item under the provisions of 10CFR50.5C(e). Our most recent interim report was logged TXX-6494, dated June 12, 1987. We are continuing our evaluation and anticipate submitting our next report by October 7, 1987. 1 Very truly yours, k $W& , W..G. Counsil RWH/gj c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) '
.tOO North Olive Street LB 8I Dallas. Texas 7201
\ .__ --________ _ ___- O
ai i F M Log # TXX-6610
.- File # 1.0110 4 L : 907.6 .
.5 3 Ref: 10CFR50.55(e) illELECTRIC wmim c. coumu Esecurare Vwe President l U. .S. Nucle? Regulatory Commission Attn: Document Control Desk ,
Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 B0P SAFETY RELATED INSTRUMENT SET P0INTS SDAR: CP-86-81 (INTERIM REPORT) Gentlemen: On November 7,1986, we verbally notified your Mr. T. Westerman of a. potentially reportable item involving the calculations for B0P safety-related instrument setpoints indicating that omissions of required data or use of
' incorrect information occurred in performing the calculations required in Reg.
Guide 1.105, Rev. 1. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Our latest . interim report logged was TXX-6271 dated February 6, 1987. The. engineering evaluation of this issue has been comoleted and the initial ,
.results indicate that this issue is not reportable pursuant to 10CFR50.55(e). ;
We are in the process of final verification of those results and anticipate comp'ation by August 21, 1987. We will submit our next report by September 18, 1987. Very truly yours, L Uh W. G. Counsil By: - M G. S. Ke: ley G / -- Manager, riuclear Licensing RDD/mlh c - Mr. R. D. Martin, Region IV , Resident Inspectors, CPSES (3) 1 na oi- s e, to si wu,. r~, mos
- _ _ _ _ _ _ _ - - --- -- 1
7 M Log # TXX-6568
== File # 10110 2
._ 963.9
=- = Ref# 10CFR50.55(e) .
TilELECTRIC vimiam c. coumii uly 17, 1987 i Decutne Vwe Prenkat 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 HANGER SPLICE WELDS SDAR: CP-86-82 (INTERIM REPORT) Gentlemen: On January 6,1987, we notified you by our let'ter logged TXX-6201 of a deficiency involving certain welds which are used to splico cable tray hanger ' (C1H) channel sections end-to-end to form posts 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 latest interim report was logged TXX-6383, dated April 3, 1987. The as-built field verification program for the Unit 1.CTH installations has ; been completed. Through the review of as-built drawings and nondestructive examination, 131 splice welds have been identified for disposition as documented by the resulting nonconformance reports. The associated construction activities for these installations are expected to be completed no later than October 30, 1987. Deficiencies identified in Unit 2 are being corrected as part of the Unit 2 ' design verification program. Currently, the Unit 2 program is expected to be completed by February 26, 1987. Our next report on this issue will be submitted no later than November 20, 1987. Very truly yours, [k W W. G. Counsil By: f-G. S. Keeley - - J'
~~
Manager,NuclearNcens!ng DAR/dl c - R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 North Ohvc Strat LB 81 Dallas, Texas 75201 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1
i M Log # TXX-6593 File # 10110
.-- 909.2
= '= Ref # 10CFR50.55(e) 'l 1UELECTRIC i William G. Cournil '
Execurne Vwe Preudent U. S. Nuclear Regulatory Commission , Attn: Document Control Desk l Washington, D. C. 20555 i
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 j STATIC COMPUTER MODEL FOR THE SAFEGUARDS BUILDING ! SDAR: CP-87-01 (INTERIM REPORT) i Gentlemen: ' On January ~14, 1987, we verbally notified your Mr. I. Barnes of a potentially reportable deficiency involving an error in the finite element computer model for the Safeguards Building. This is an interim report under the provisions l of10CFR50.55(e). The error involves a fixed support point at the roof elevation where no such support exists. This condition produces an incorrect load distribution in the computer model, which will require a reevaluation'1 of the structural members in the building. We are continuing our evaluation as a part of the Design Basis Consolidation Program (DBCP). Although our preliminary review has indicated that no condition adverse to.the safety of plant operation has resulted from this item,.the portion of the DBCP which includes this deficiercy is currently imcomplete. Our next report on this issue will be submitted no later than September 3, 1987. Very truly yours, i M W. G. Counsil By: G. S. Keeley
/
/,
.MCP/gj Manager,Nucleart.ysing c - Mr. R. D. Martin, Region IV l Resident Inspectors, CPSES (3) s I
500 North Ohve Street LB 81 Dallas. Texas 75.'01 l i
l M Log # TXX-6578 File # 10110 ! l
-._ .-- 903.11
= C Ref # 10CFR50.55(e)
TilELECTRIC 4 wmim c. coumu July 21,1987 Esucutwe Vwe Preudem U. S. Nuclear Regulatory Commission ATTN: Document Control Desk , Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOLKET NOS. 50-445 AND 50-446 ENVIRONMENTAL QUALIFICATION OF CABLE TO POST-LOCA HIGH RANGE RADIATION MONITOR SDAR:~ CP-87-05 (INTERIM REPORT) Gentlemen: On April 6,1987, we verbally notified your Mr. C. Hale of a deficiency involving the failure of High Range Radiation Monitor (HRRM) coaxial cables to j meet the minimum cable environmental qualification for dielectric insulation requirements. This is an interim report of a potentially reportable item 1 under.'the provisions of 10CFR50.55(e). Our most recent interim report was logged TXX-6430, dated May 6, 1987. As noted in our previous report, we have reviewed the Sorrento Electronics
.(f ) 10CFR21 Notification and have verified that it-is applicable to CPSES l ew aipment. We also advised that our evaluation of this issue would be i contingent upon the recommendation of corrective actions by SE.
SE has recommended possible solutions for our consideration. Currently, we are evaluating the possible solutions given by SE. The results of this evaluation are currently scheduled for completion by October 30, 1987. We anticipate submitting our next report on this issue no later than November 20, 1587. Very tr 1 yours, W. G. C unsil By: , . G. S. Keeley . - Manager, Nuclear Licqn/ing c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 North Olive Street LB 81 Dallas, Texas 73201 j
L,. M Log # I'/X-6574
'F E File # 10110
. 907.0 t: ,
= = Ref: _10CFR50.55(e) 1UELECTRIC Williarn G. Counsil '
f secut*Ve Vrce Prnuknt 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.-03 (INTERIM REPORT) Gentlemen: On April 20, 1987 we verbally notified your Mr. Cliff Hale of a deficiency regarding the accuracy tolercnce of calibration for- a Dead Weight Tester used as a standard in calibrating pressure instrumentation. This tester has been used in the Operations (Instrumentation & Controls) test / maintenance program. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Our latest report was logged TXX-6456, dated May 20, 1987. We have reviewed all calibration standards used in the Operations (Instruments Controls) Program and have identified the following standards which do not meet the required accuracy tolerances. MIX TAG ! EESGRIPTION .TlPE 10-0172 Multimeter Calibrator Temperature IC-0180 Pneumatic Dead Weight Tester n usure IC-0181 Dead Weight Tester Prassure IC-0182 Dead Weight Tester . Pressure IC-0183 Pneumatic Dead Weight Tester Pressure 1 10-0280 Digital Pressure Indicator Pressure IC-2304 U-Tube Manometer Pressure Currently, we are evaluating the effect of these inaccuracies on programs important to safety (for example, cold hydrostatic tests, preoparational tests,etc.). 400 North Olive Street LR 81 Dallas, Tens 75201 l a
- d- t TXX-6574 !
July 21, 1987 Page 2 of 2 ) j The above evaluation is scheduled for completion by September 30, 1987. We f will submit our next report on this issue by October 23, 1987. 4 Very truly yours, bb'. X l'hlt W. G. Counsil i By: h '/ G. S. Keeley / Manager, Nucl @ g ensing ! BSD/mlh c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) l i i J
- F==ma Log # TXX-6619
""" E File # 10110
= .- .908.3
= = Ref: 10CFR50.55(e) 1UELECTRIC wmim c. counsit August 7, 1987 E ecutive %ce Prestdent 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 ! TERMINAL STUDS IN PK-2 TEST BLOCKS I SDAR: CP-87-09 (FINAL REPORT) i Gentlemen: On May 7,1987, we verbally notified your Mr. Cliff Hale of a deficiency 1 involving terminal studs in PK-2 test blocks. We have' submitted an interim report logged TXX-6495, dated June 5, 1987. Visual inspection of the studs used in PK-2 test blocks was performed and no defects were observed. A review of the applicable elementary diagrams was also performed which determined that the test blocks were located on test circuit for the 6.9kV switchgear and not on operating circuit. Failure of a PK-2.. test block as installed in relay circuitry.at CPSES would not cause equipment or system failure. Ac a conservative measure, the existing studs in PK-2 test blocks are being replaced as recommended by the vendor (General Electric). The replacement studs will be manufactured with the proper material and will also receive the requ'. red stress relief. i In the event this deficiency had remained undetected, no condition adverse to t'ne safety of plant operations would exist. Therefore, this deficiency is not reportabic under the provisions of 10CFR50.55(e). Records supporting this cor:lusion are available for your inspectors to review at the CPSES project j site. ' L Very truly yours, M W. G. Counsil ! WJH/mlh c - Mr. R. D. Martin, Region IV i Resident Inspectors, CPSES (3) { 400 North Olive Street LB 81 Dallas, Teus 75201 l
\
t
M Log # TXX-6617 F9 .-- File # 10110
. 906.2 r = Ref # 10CFR50.55(e).
nlELECTRIC \ wmu a c3w August 5, 1987 Eueutwe %ce Presulent U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) l DOCKET NOS. 50-445 AND 50-446 CONTAINMENT P/T ANALYSIS COMPUTER ERROR SDAR: CP-87-12 (INTERIM REPORT) Gentlemen: i On June 12, 1987, we notified you by our letter logged TXX-6512, of a deficiency involving a containment pressure / temperature analysis computer error. This is an interim report of a deficiency which is reportable under the provisions of 10CFR50.55(e). Our reanalysis of the Main Steam Line Break accident has been completed. The results of this analysis predict a containment peak temperature of 344.50F, compared to the original value of 3330F used in the design basis. This : evaluation is scheduled to be completed on September 30, 1987. We will submit our next report by December 11, 1987. Very truly yours, Y,(,(canik W. G. Counsil By: C pK S. Marshall l Supervisor, Generic Licer. sing ! JCH/mgt ' i l c - Mr. R. D. Martin, Region IV , Resident Inspectors, CPSES (3) 400 North O!ive Street LB 88 Dallas, Texas 73201 I ll I
l M Log # TXX-6632
== File # 10110 r
d 907
= Ref # 10CFR50.55(e)
TUELECTRIC William G. Counsil ' Esecurove %cc Presodent 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 LIMIT SWITCH. WIRING
,SDAR:
CP-87-16 (INTERIM REPORT) Gentlemen: On June 1, 1987, we verbally notified your Mr. H. S. Phillips of a deficiancy ; involving conduit and cable which feed safety related valve limit switches and which were not installed in accordance with the wiring drawings. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). We'are continuing our evaluation and anticipate submitting our next report by October'7, 1987. Very truly yours V)Q W. G. Counsil JCH/gj c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) { 400 Nonh Olive Street LB 81 Dallas. Tew 73201
M Log # TXX-6554 F ""- File # 10110
.-- 920.1
= = Ref # 10CFR50.55(e) 1UELECTRIC wunam c. counsu Encutne Voce Pressdent ,
July 8* 1987 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET N05. 50-445 AND 50-446 VALIDYNE 15 VOLT DC POWER SUPPLY ( SDAR: CP-87-17 (INTERIM REPORT) Ge'ntlemen: On June 8,1987, we verbally notified your Mr. Ian Barnes of a deficiency in the operation of power supplies manufactured by Validyne Engineering Corporation. The power supply contains a 15V DC regulator which can exhibit a high frequency oscillation under certain load conditions. This is an interim report of a potentially reportable item under the provisions 'of 10CFR50.55(e). Our evaluation of this issue is continuing. We expect to submit our next report by August 7, 1987. Very truly yours,
', tv b W. G.,Counsil By:
~
G. S. Keeley C ' /- Manager, Nuclear / censing
]
RSB/mgt c - R. D. Martin, Regia IV l Resident Inspectors, CPSES (3) L I l l .n - o,. s- a ,, v.,,.. r - ma,
- _ _ _ _ - - - - - - - _ _ - - - ------------ ---- - A
i i C- Log #'TXX-6626
- t. == File # 10110 I
! h j 907
= = Ref: 10CFR50.55(e) 1UELECTRIC wmm c. coon,a Augud 7, 1987 Esecutne Vwe Prmtent U. S. Nuclear Regulatory Conmission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE PEAX STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 VALIDYNE 15 VOLT DC POWER SUPPLY ' SDAR: CP-87-17 (FINAL REPORT) Gentlemen:. On June 8,1987, we verbally notified your Mr. Ian Barnes of a deficiency in-the operation of power supplies manufactured by Validyne Engineering Corporation. The power supply contains a 15V DC regulator which can exhibit a i high frequency oscillation under certain load conditions. This oscillation would result in' a minor zero shift in output of some of the signal conditioning cards. Calibration accuracy is compromised by less than 0.1% by the presence of this oscillation. We have evaluated this issue for safety significa'nce and have determined the power supplies which contain the subject regulator are installed at CPSES in a singular system, the Emergency Response Facility (ERF) computer system. As installed, were the deficiency to have rcmained uncorrected it would not have had an adverse effect on the safety of the plant operations. We have confirmed the specific defective component is not used in a safety-related application. We have therefore concluded that this issue is not reportable under the provisions of 10CFR50.55(e). Records supporting this determination are available for your inspectors to review at the CPSES site. i Very truly your , !
<i p W. G. Counsil i RSB/mlh '
c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3)
/
r 400 North Obve Street LB 81 Dallas, Tens 73201 __.___-_-____b
L v ! l' , ,Q
-N Log # TYY-6607 i FE .
File # 10;;0 u9.? i
= r Ref # 10CFT<50.55(e) !
7 1UELECTRIC l l
, ,g . August 3, 1987 uane na treau 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 COMP 0NENT'C00 LING WATER (CCW) HEAT EXCHANGER SEISMIC QUALIFICATION SDAR: CP-87-18 (INTERIM REPORT) p Gentlemen: On June' 8,1987, we verbally notified your Mr. I. Barnes of a deficiency ~ involving errors identified during a review of. the. vendor qualification documentation for the.CCW heat exchangers that indicated the seismic and nozzle loading conditions exceed specification requirements'and could be overstressed. This is an interim report of a potentially reportable item under the provisions of 10CFR 50.55(e).. Our initial evaluation of the subject heat exchangers showed stresr9s exceeding the allowables in' both-the. heat exchanger shell and the supports. We have also determined that the specification nozzle loadings contributed a significant portion of the total heat exchanger loading. To reduce the conservatism in the initial evaluation, the heat exchangers were evaluated further using the as-built piping' loadings. Piping as-built nozzle loadings were obtained and were found to be significantly lower than the specification loadings. The nozzle loads and the seismic inertia loads were applied to a finite element model (FEM) of the heat exchanger to obtain shell loadings, support loadings, and reaction forces. Overall heat exchanger shell stresses, local shell stresses at nozzles, local shell stresses at supports, and support stresses have been analyzed. l 300 Nonh Ohve Street LB 81 Dallas, Tetas 73201
l n
.,.J TXX-6607 i August 3, 1987' Page 2 of 2 Preliminary results from this evaluation indicate that the Unit I heat exchangers meet ASME Boiler & Pressure Vessel code and FSAR stress limits.
Before the evaluation is finalized, the following items must be addressed:
- 1) The first mode frequency of the heat exchanger was determined to be less than 33 Hz. The impact of the heat exchanger flexibility on the piping analyses must be determined.
- 2) The heat exchanger anchorage loadings must be evaluated.
- 3) The heat exchanger concrete pedestal supports must be evaluated to determine the pedestal natural frequencies. If the pedestal <
fundamental frequency is found to be less than 33 Hz, further evaluation of. the heat exchangers will be required.
- 4) Unit 2 as-built nozzle loadings must be obtained b'efore an evaluation can be performed on the Unit 2 heat exch' anger. ;
- 5) Piping mode shapes and frequencies for the piping connected to the CCW heat exchanger nozzles must be obtainad to justify combining..
the four nozzle seismic loadings as performed in the evaluation. Our next report on this issue will be submitted no later than September 11, 1987. Very truly yours, it' W. G. Counsil MCP/mgt c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) ,! i l 1 1 I i k I
~
' (
i l l M Log # TXX-6558 {
== File # 10110 !
h
=
5
=
909.2 Ref # 10CFR50.55(e) TUELECTRIC ws.n c. counsii July 8,1987 l Esecutove Vwe Presukat U. S. Nuckar Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE FEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 COMP 0NENT COOLING WATER (CCW) HEAT EXCHANGER SEISMIC QUALIFICATION SDAR: CP-87-18 (INTERIM REPORT) Gentlemen: On June 8,1987, we verbally notified your Mr. Ian Barnes of a deficiency involving errors identified during review of vendor qualification ' documentation for the CCW heat exchangers which requires additional analyses to be performed. Preliminary results indicate that the error involves the point of application of the nozzle loads by the vendor (Struthers Wells) and requires a reevaluation of the equipment qualification for the CCW heat < exchangers. This is an interim report of a potentially reportable item under the provisiors of 10CFR50.55(e). A complete description of the planned evaluation is currently being proposed and will be submitted no later than August 5,1987. Very truly yours,
,%., o n<,lt W. G. Counsil M
By: / G. S. Keeley e f Manager, Nuclear Lice,nsing MCP/mgt c - R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 North Olive Street LB 81 Dallas, Tens 7:20I _ _ _ - _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ - _ - - _ - - -- _ - - _ - _ - - - - - _ - --- ^
%I M Log # TXX-6557 F "-- File # 10110
.-- 910.3
= = Ref # 10CFR50.55(e) 1UELECTRIC WHHam G. Cou Al '
Emvtive Vwe Presulers: 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 ! AMBIENT TEMPERATURE EFFECTS ON MSIV ACTUATORS SDAR: CP-87-19 (INTERIM REPORT) Gentlemen: On June 11, 1987, we verbally notified your Mr. Bob Warnick of a deficiency involving the ambient' temperature effects on MSIV actuators. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). The manufacturer (Rockwell International) has advised TU Electric that the thermal compensating accumulator on the valve actuator attached to the Main Steam Isolation Valves (MSIV) may not function in accordance with the CPSES procurement specification. The specification requires the actuator to function correctly through an ambient temperature swing of 800F. Test results', recently acquired by Rockwell, indicate the greatest possible ambient swing, which will not impair the function of the actuator, is 340F. At present, any increase above the 340F ambient temperature swing could result in excessively high hydraulic pressure to the unit which could damage components in the actuator or possibly inhibit the main pump hydraulic solenoid valves from shifting. In the latter case, the actuator would not perform the required safety functions. These adverse pressures should occur only during increasing thermal transients of the ambient environment while the actuator is in the open position. This could typically occur during a plant heatup with the valve and actuator open. This condition is applicable to both Unit 1 and 2 and affects the main steam isolation. valves and associated A-180 actuator and accumulators. Specific tag numbcrs for the affected valve are as follows: 1-HV-2333A, 1-HV-2334A, 1-HV-2335A, 1-HV-2336A 2-HV-2333A, 2-HV-2334A, 1-HV-2335A, 2-HV-2336A 400 Nonh Olive Street LB 81 Dallas, Texas 73201
/
, c.
r TXX-6557 ' July 10,196'l Page 2 of 2 ! Our evaluation includes the following:
- 1) The Mechanical Specification (MS-76) currently specifies an 800F ambient temperature (40-1200F) increase. This requirement will be reviewed to determine if the temperature increase can be limited to 340F.
- 2) The allowable temperature increase can be raised to 550F by increasing the nitrogen gas precharge pressure to 90% of the hydraulic operating pressure of 2650 psi instead of the current precharge of 1325 psi. The system review will assess if the temperature transient can be limited to a rise of 550F.
3)' If thr. review indicates that the temperature gradient cannot be limited to one of the lower values suggested above, the A-180 actuators will be modified so that the original specification requirement of 800F can be achieved. The modification may consist of the installation of a pressure relief valve in the hydraulic system for each valve.
- 4) Equipment environmental and sel:mic qualifications.
The results of the above evaluation will be used to assess the significance of this issue. Our next report on this issue will oe submitted no later than September 18, 1987. l Very truly yours, nd W. G. Counsil ! By: d, . , 6 / G. S. Keeley Ns Manager, Nuclear Li Vnsing c - R. D. Martin, Region IV Resident Inspectors, CPSES (3) i l 4 u_______m.__ . _ _ - _ . _ _ _ _
b \; r M Log # TXX-6576 FE File # 10110 908.3 [ r = Ref # 10CFR50.55(e) 1UELECTRIC I
. wmiam c. counsil Executne Vu:e Presulent July 17, 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 CRACKED PORCELAIN CONNECTORS ON 6.9 KV TRANSFORMERS SDAR: CP-87-20 (INTERIM REPORT) Gentlemen: On June 17, 1987, we verbally notified your Mr. S. H. Phillips of a deficiency involving cracking identified in the porcelain bus connectors for 6.9 kV transformers. Investigation determined the cracks to be due to overtorquing of the connection hardware. This deficiency could possibly result in insulator failure during a seismic event. This is'an interim report of a potentially reportable item under the provisions of ~10CFR50.55(e). Our evaluation of this issue is continuing. We expect to submit our next report by August 28, 1987. Very truly yours, W. G. Counsi i By: - , G. S. Keeley
/
cs L Manager,NuclearLicenyng WJH/wjh 1 c - Mr. R. D. Martin, Region IV ' Resident Inspectors, CPSES (3) i
)
400 Nonh OI ve Street LB St DaBas, Tens 7201
M Loa ' FM Fiie#TXX-6569
# 10110
.-- .. 908.3
= = Ref: 10CFR50.55(e)
. n/ ELECTRIC Williarn G. Counsil
Execuene Vne Premient I
' U.. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
. COMANCHE PEAK STEAM ELECTRIC. STATION (CPSES) i DOCKET NOS. 50-445 AND 50-446 EFFECTS OF THERM 0 LAG ON CABLE DERATING FACTORS ,
.SDAR: CP-87-21-(INTERIM REPORT) !
Gentlemen: On June 17, 1987, we verbally notified your Mr. H. S. Phillips of a ' deficiency involving recent evaluations. establishing derating factors.of 31% for single cable trays and 21% for single conduits enclosed in thermolag. These derating factors are larger than the derating factor (10%) used in the initial cablei sizing calculations. Thir derating factor could result in-cable degradation if cables are operated in ~ excess of the wrapped derated cable capacity. This is an interim report of. a potentially reportable item under the provisions of 10CFR50.55(e). Our evaluation of this issue is continuing. We expect to submit our next report by. August 17, 1987. Very truly yours, i Ohl b i W. G. Counsil By: r ; G. S. Keelt f / -- - Manager,NucleaQ4 censing ! JDS/mlh c - Mr. R. D. Martin, Region IV-l Resident Inspectors, CPSES-(3) l 4 400 North Olive Street LB 81 Dallas. Texas 75201 L_ ___ _ U
Log # TXX-6579
======== Fi1e # 10110 FE 2 .--
903.8 Ref # 10CFR50.55(e)
= =
1UELECTRIC July 17,1987 William G. Coumil j Esecutne Vwe Preudent l U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Warhington, D.C. 20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NO. 50-445 LINER ATTACHMENT MATERIAL REQUIREMENTS: IMPACT TESTING ! SDAR: CP-87-22 (INTERIM REPORT) Gentlemen: On June 17, 1987, we verbally notified your Mr. Shanncn Phillips of a deficiency involving FSAR and specification requirements regarding impact testing of materials attached to the containment liner. Specifically, Section 3.8.1 of tke q 3 FSAR and Specification 2323-SS-14 require notch toughness impact tests for i materials attached to the containment liner. One 3/4" plate, to wliich pipe j support CT-1-031-010-C92S is installed, is attached to a 1" containment liner ! insert plate. The Mill Test Report for the 3/4" plate did not contain the results 1 of the notch toughness impact test. This is an interim report of a potentially 1 reportable item under the provisions of 10CFR50.55(e). Currently, an engineering evaluation of the materials and weld configurations for ! the attachment to the liner plate is in process. lhis evaluation is required to ; determine any safety significance and reportablity pursuant to 10CFR50.55(e). We will provide our next report by August 31, 1987. Very truly yours, !
$h, tw W. G. Counsil '
/ '
By: A '.
~
G. S. Keeley C
/
[ Manager,NuclearLic/nsing BSD/amb c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) ) i 400 Nonh Olive Street LB 81 Dallas, Texas 752OI
.. I
(v ' Nb ;e 1 M Log # TXX-6565 t == File # 10110 h,,,,, 5:: 903.9 Ref # 10CFR50.55(e) TUELECTRIC wunun c. coumii uly 27,1987 Esecuent Hct Permtent c U.-S. Nuclear Regtalatory Commission ATTN: Document Control Desk , Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 CONDUIT UNIONS SDAR: CP-87-23 (FINAL REPORT) Gentlemen: I On June 26, 1987, we verbally notified your Mr. Bob Warnick of a deficiency involving not securely Class IE conduit installation runs coupled with unions which were tightened. under the provisions of 10CFR50.55(e).We have information The required determined-that this condition! follows. DESCRIPTION' During random validation inspections, couplings were observed loose in both safety and non-safety class conduit installations. initiated Corrective Action Request (CAR 87-002), the root cause of the In evaluation of the site deficiency was determined'to be inadequacies in the associated. construction and inspection procedures. Specifically, these procedures omitted provisions for the verification of secured conduit couplings. The basis of this weakness in the procedures is speculative, however, the intent of the procedures infers that " good workmanship practice" would' preclude the need to formally address securing the conduit couplings. SAFETY IMPLICATIONS i In the event conduits separate as a result of loose couplings due to equipment vibration or during a seismic event, the two conduit sections would be held in a cantilever position by the respective supports. Continued movement of the free conduit end could damage the enclosed cable which could result in the inability plant of associated operation safety-related systems to perform as required for and shutdown. CORBECTIVE ACTICN A 100 percent validation program is being performed by Ebasco under Procedure TE-FVM-CS-033, " Design Control of Electrical Conduit Raceways for Unit 1-Installation in Unit I and Common Areas". This program includes a field
- c. .
400 North Ohse Street LB 81 Dallas, Tens 7%I 1 c_____1___________----_-----
- m. 0 l
TXX-6565 July 27, 1987 Page 2 of 2 1 valid'ation of the . Unit 1 Train A and B, and non-safety related (over 2 inches in diameter conduit installations. Also this program includes a validation of the neo-w) eld, communications, security system conduits great inches in diameter and safety-related radiation monitoring-system conduits. j i
.Non-safety related conduit 2 inches and under in diameter are not included within the scope of the issue. Loose conduit unions found during the walkdown i will be identified and dispositioned using deficiency documents. The I projected completion date for identifying and correcting any loose unions is December 1, 1987.
Unions installed in Unit 2 conduit are noted on the conduit isometrics drawings. . Conduits having unions will be identified by review of the isometric drawings and the unions will be inspected. - As applicable, correction of. the identified loose unions will be dispositioned using i deficiency' 1987. documents and the completion date of activities is December 1, The following Units 1.& 2 installation and inspection procedure / specification wil1 be revised by September 4, 1987 to require validation of conduit union connections for tightness: ! ECP " Exposed Conduit / Junction Box and Hanger Fabrication and Installation", ',
'2323-ES-100 " Electrical Erection Specification".
Records supporting our evaluation will be available for your inspectors review at the CPSES site after December 15, 1987. s i Very truly yours, b th;$. W. G. Counsil
/ J By: MM < b G. S. Keeley '
Manager, Nuclear Licensing
/~ !
0AR/mgt i c - R. D. Martin, Region IV ) Resident Inspectors, CPSES (3)
.w..
i
^
. _ _ _ . _.____________ ____ d
M Log #.TXX-6587 F -- File # 10110 i
= ,_. 906.2 r r Ref: 10CFR50.55(e) 1UELECTRIC
}
1 wmiam c. coumu becutor Het Ptnunt July 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 ! CONTAINMENT LINER NELSON STUD BENDING SDAR: CP-87-24 (INTERIM REPORT) Gentlemen: On June 29, 1987, we verbally notified your Mr. H. S. Phillips' of a deficiency involving engineering approval provided to allow bending of Nelson stud anchors on the containment liner.in excess of FSAR requirements. This is-an i interim report of a potentially reportable item under the provisions of ' 10CFR50.55(e). We are continuing our evaluation and anticipate submitting our next report on this issue no later than August 28, 1987. Very truly yours, W. G. Counsil i By: ' t/ G. S. Keeley JDS/mlh Manager, N censing c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) I I 0 I 400 North Ohve Street LB 81 Dadas. Tens 73201 __________-_ A
i s Pommm. Log # TXX-6609
== File # 10110 L
._ d 909.5
= = Ref. # 10CFR50.55(e) 1UELECTRIC
, July 29,.1987 William G. Counsil Execursve %ce Presutent U. S. Nuclear Regulatory Commission t
Attn: Document Control Desk' . Washington, D. C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND.50-446'
~DG FUEL OIL TANK VENT MISSILE PROTECTION:
SDAR: CP-87-25 (INTERIM REPORT) Gentlemen: i On June 29, 1987, we verbally notified you'r Mr. Shannon Phillips of a deficiency involving failure to provide missile ;,cotection for the diesel fuel oil storage tank vent piping. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Specifically, during the development of the design basis document for the diesel fuel oil storage system, it was' noted that the vent piping for the fuel oil storage and day tanks were not afforded missile protection. Additionally, portions of. the vent' piping were determined 'not to be seismically designed as required. Our evaluation for possible adverse effects on plant safety and. deportability under the provisions of.10CFR50.55(e) will be performed concurrently with engineering evaluations for the required corrective actions.- 3 We anticipate submitting our next report by September 15, 1987. Very truly yours, j ffz h4MJ b W. G. Counsil By: G. S. Keeley c - . L! Manager, Nuclear Lic/ensing BSD/gj c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 Norm O!rve Street LB BI Dallas. Texas 120I
i-M
" " " ' " " " Log .# TXX-6594 File # 10110
. .-- 909.5
= =- Ref # 10CFR50.55(e) ;
1UELECTRIC \ wmiam c. counsu July 28, 1987 4 Esecutne %ce Presskat i 9- V. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 DIESEL GENERATOR FUEL OIL TRANSFER PUMP SUCTION LIFT SDAR: CP-87-26 (INTERIM REPORT) Gentlemen: On June 29, 1987, we verbally notified your Mr. H. S. Phillips of a deficiency involving the die.sel generator fuel oil transfer pump. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). Specifically, recent calculations performed under the Design Validation Program (DVP) indicated that the installed piping configuration for " worst case" operating conditions requires a pump suction lift greater than r originally expected as required by the diesel generator fuel oil transfer pump > specification. Results of 'special testing ' conducted subsequent to these calculations are currently being evaluated and are expected to confirm the adequacy of the installation without modification. Preliminary results of this evaluation ; indicate'no physical changes to the transfer system will be required. We are continuing our evaluation as a part of the Design Validation Program and anticipate submitting our next report by August 29, 1987. I Very truly yours, f l l W. G. Counsil MCP/gj ! i c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) i 400 Nutth Olive Street LB 81 Dallas. Texas 73201 l
l 1 M Log # TXX-6583 j F9 File # 10110 909.2 ! r -- Ref: 10CFR50.55(e) i 1UELECTRIC William G. Counsil ' becutive Vwe Vrewlent (J . S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 l 1
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) i DOCKET N05. 50-445 AND 50-446 GALVANIC CORROSION IN CCW AND SW SYSTEMS SDAR: IP-87-27 (INTERIM REPORT) Gentlemen:
)
On June 29, 1987, we verbally notified your Mr. Shannon Phillips of a potentially reportable item involving galvanic corrosion in Component Cooling Water (CCW) and Service Water (SW) systems. This is an interim report under { the provisions of 10CFR50.55 Original notification included the CCW System, this was an error as(e). only the SW system is involved. ] { s During design verification of the flow element specifications, a potential design discrepancy has been identified involving dissimilar metals in the SW system. The annubars installed in this system are made of monel material. The piping material is both carbon steel and stainless steel. Bolted contact of the annubar mounting flange to the process flange on the piping could result in galvanic corrosion and possible degradation of the process flange. I This conditions could render the annubar inoperable and/or violate the integrity of the piping systems. Currently, we are reviewing this system to identify the components affected by this condition. Our next report on this issue will be submitted no later than August 21, 1987. Very truly yours, Y.X, W. G. Counsil By: G. S. Keeley C / MCP/mlh Manager, Nuclear JMcensing c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) ; l l 400 Nonh Ohve Street LB 81 D.allas, Texas 73201 l U
M Log # TXX-6585 llll"."" 9 File # 10110
.-- 910.4 r = Ref: 10CFR50.55(e) 1UELECTRIC wimam c. counsa uly 28, 1987 Enesutne Vwe Preudent U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-415 AND 50-446 . CONDENSATE STORAGE TANK OVERPRESSURIZATION SDAR: CP-87-28 (INTERIM REPORT)
' Gentlemen:
On' June 29, 1987, we verbally notified your Mr. Shannon Phillips of a potentially reportable item involving the capacity of the condensate storage tank.(CST). This is-an interim report under the provisions of 10CFR50.55(e). During the Design Bases Consolidation Program, a potential problem related to
.the capacity of the CST was identified. Specifically, the maximum permissible normal operating level established by design may not leave sufficient tank
-volume to accommodate. excess condensate during certain plant transients.
Water addition to the CST during these transients may exceed the tank overflow capacity resulting in overpressurization and potential rupture of the tank. A tank rupture would. adversely impact the operability of the auxiliary feedwater system. We are continuing cur evaluation to determine if operational changes to the
. CST normal level can accommodate expected transients and if a tank rupture was a credible event with the present design. We will submit our next report by September 21, 1987. >
Very truly yours, ik W. G. Counsil By: A / MCP/mlh G. S. Keeley e Manager, Nuclear ~ing s c - Mr. R. D. Martin, Region IV Resident' Inspectors, CPSES (3) i l i
.100 North Ohve Street LB 81 Dallas. Texas 7320!
______o
= N== Log # TXX-6608 FE 1 -
File # 10110 _ 909.2
= = Ref # 10CFR50.55(e) 1UELECTRIC wmm c. coumit July 29, 1987 Esecuene Vwe Preudent 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 COMPANENT COOLING WATER (CCW) ISOLATION FOLLOWING REACTOR COOLANT PUMP (RCP) THERMAL BARRIER RUPTURE SDAR: CP-87-29 (INTERIM REPORT) Gentlemen: On June 29, 1987 we verbally notified your Mr. Shannon Phillips of a deficiency involving instrumentation and controls utilized for isolation of Component Cooling Water (CCW) to the Reactor Coolant . Pump (RCP) thermal barrier heat exchanger. This is an interim report of a potentially reportable item under the provisions of 10CFR50.05(e). Specificaih,a tube failure in the RCP heat exchanger is isolated by a ,non Class 1E Temperature actuated isolation valve. Backup isolation capability for this valve is provided by operator action using the. containment isolation valves. To take credit for operator action as the-backup isolation capability, the instruments used to detect the break must be Class IE (required by RG 1.97' Instrumentation.for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident). Currently, instrumentation capable of detecting a thermal barrier tube leak , (flow, temperature, radiation monitors, and CCW surge tank high level), are j non Class IE. , Our evaluation for possible adverse effects on plant safety and deportability under the provisions of 10CFR50.55(e) will be performed concurrently with engineering reviews required to resolve the deficiency. We anticipate submitting our next report on this issue no later than September 15, 1987. Very truly yours, W. G. o nsi W j-By :_, G. S. Keeley e / Manager, Nuclear Licin~ sing MCP/gj V c - Mr. R. D. Martin, Region IV Resident Inspectorg,,CjSES g {32, ta si carias. Texas 7520i
M Log # TXX-6601
'F E m File # 10110 903.5 r r Ref. # 10CFR50.55(e) !
illELECTRIC Williarn G. Counsil ' Esecuuve %ce hrsakar 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 ROLL-AWAY' MISSILE SHIELDS (RAMS)- ; SDAR: CP.-87-30 (INTERIM REPORT) )
' Gentlemen:
' On July 1,1987, we verbally notified your Mr. Shannon Phillips of a deficiency involving errors in the seismic qualification report supplied by the vendor of the Roll-Away Missile Shields (RAMS). A preliminary review of these errors indicated that the RAMS were overstressed for the designed seismic conditions detailed in the_ applicable specifications. . This is an ' interim report of a potentially
. reportable item under the provisions of 10CFR50.55(e).
Recently. performed calculations,. based on vendor supplied as-built ' drawings, . indicated that these errors were not safety significant. However, subsequent field inspections identified discrepancies'in the vendor as-built drawings. Additionally, it has been determined that the current configuration of the refueling crane rails used by the RAMS is not seismically' qualified for that application. It is anticipated.that modifications to prevent transfer of lateral
,(RAMS) loads to the crane rails will be required.
We are currently evaluating these deficiencies for reportebility under the
- provisions of 10CFR50.55(e). We will submit our next report by October 16, 1987.
Very truly yours, 7b> W. G. Counsil I BSD/amb c .Mr. R. D. Martin, Region IV . Resident Inspectors, CPSES (3) ' 400 Nonh Ohve Street LB 8b Dallas, Teus 75201 _____-_-___.m_-_ - __-
l N Log # TXX-6620 F9_ .-- File # 10110 909.5
= r Ref # 10CFR50.55(e) illELECTRIC William G. Counsil * '
l becuine he Presdens l U. S. Nuclear Regulatory Commission
. Attn: Document Control Desk Washington, D. C. 20555 l
I
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 DIESEL GENERATOR CONTROL AIR REGULATOR ! SDAR: CP-87-31 (FINAL REPORT) ! Jentiemen: 'I On July 6,1987, we verbally notified your Mr. H. S. Phillips of a potential deficiency involving the air pressure regulator on the IMO Delaval Diesel j Generators 'which may contain manufacturing defects in the dripwell gasket i seating surfaces. These defects could cause a loss of control air which in turn could cause a loss of starting air pressure. 1 The 10 CFR Part 21 report submitted by IM0 Delaval indicates the defective regulators are identifiable by the date of manufacture stamped on the' bottom of the regulator. It has been confirmed that none of the defective regulators , (including four installed regulators and three spares) are in use at Comanche Peak Steam. Electric Station. ~ As discussed above, no deficiency existed at Comanche Peak Steam Electric Station, therefore no report pursuant to.10CFR50.55(e) is required.
-l Very truly yours, V). N r .a ,. .:
W. G. Counsil i By:
- 0. S. Marshall !
Supervisor, Generic Licensing MCP/gj c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) 400 Nors' Olne Street LB 81 Dallas, Texas 75201 i
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= N=mm- Log # TXX-6624 i F-- File # 10110
._ .- 917.1 l r = Ref # 10CFR50.55(e) 1UELECTRIC muiam c. counsa August 7, 1987 Emwne Vue PreuJent U. S. Nuclear Regulatory Commission i Attn: Document Control Desk Washington, D. C. 20555 i
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 QUALITY CONTROL INSPECTOR QUALIFICATION SDAR CP-87-32 (INTERIM REPORT) Gentlemen: On July;8,1987, we verbally notified your Mr. R. F. Warnick of a deficiency involving apparent weaknesses in our Quality Control Inspector Qualification - Program. Specifically, procedures for establishing QC Inspector qualification criteria' did not address regulatory requirements such as the removal of unqualified personnel, the reevaluation of test personnel, certification data, physical characteristics examinations, and Level III capabilities. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). We are continuing our evaluation and anticipate submitting our next report by October 7, 1987. Very truly yours,
^
?
24 W. G. Counsil i DAR/gj ' c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) i I 1
.mo North Ohee Street LB 81 Dadas. Texas 75201
=N==- Log # TXX-6629 F " "-- File # 10110 '
.-- 910.4 r- = Ref # 10CFR50.55(e) 1UELECTRIC '
i wmim c. counsii August 7, 1987 E.tecutive %ce FreuJent U. S. Nuclear Regulatory Commission ' Attn: Document Control Desk Washington, D. C. 20555' 4
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 AUXILIARY FEEDWATER PUMP MOTOR FANS SDAR: CP-87-33 (INTERIM REPORT) Gentlemen: On July 8,,1987, we verbally notified your Mr. R. F. Warnick of a deficiency involving the rotational direction of the auxiliary feedwater pump motor cooling fans for Units 1 and 2. This is an interim report of a potentially reportable item under the provisions of 10CFR50.55(e). During startup examination of the Unit 2 auxiliary feedwater pump assembly, the rotational direction of the motor cooling fans was determined to be opposite from the design requirements. Similar condition exists in the Unit 1 auxiliary feedwater pump motor cooling fans. Further evaluation is required to determine the impact of these conditions. We will submit the next report by October 7, 1987. Very truly yours, , it' W. G. Counsil MCP/gj c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) I l I 400 North Oltve Street LB 81 Dallas Teus 73201 L_-_-___-__-_____---__--__-______
Y l
- lllll" = Log # TXX-6630 j File # 10110 1 903.9 Ref: 10CFR50.55(e) illELECTRIC wimm c. couma August 10, 1987 Ew:utove %ce Vrrudent 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 EQUIPMENT N0ZZLE LOAD INTERFACES SDAR: CP-87-34 (INTERIM REPORT) Gentlemen:
'On July 10, 1987, we verbally notified your Mr. H. S. Phillips of a deficiency involving equipment nozzle load interfaces for Units 1 and 2. This i: an interim report of a potentially reportable item under the provisions of -
10CFR50.55(e). During evaluation of the loads generated at equipment nozzles in the pipe-stress analysis, the proper interface may not have been consistently identified between the pipe stress analysis organization (SWEC-PSAS) and the equipment supplier (Westinghouse).
.Further evaluation is required to determine the impact of these conditions.
We will submit the next report by October '30,1987. Very truly yours, l ld.C. Q l W. G. Counsil By: .b Jbhn W. Beck ! Vice President i Nuclear Engineering RSB/mlh { c - Mr. R. D. Martin, Region IV I Resident Inspectors, CPSES (3) l 400 North Olive Street LB 81 Dallas. Texas 75:01 1 I ________-----_--__--J
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-~ Log # TXX-6611
-- =-
File # 10110
= .-- 903.11
= = Ref: 10CFR50.55(e) 1UELECTRIC wimam c. counsit August 7, 1987 Emutne %ce President U. S. Nuclear Regulatory Commission Attn: Document Control Desk i Washington, D. C. 20555 !
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) DOCKET NOS. 50-445 AND 50-446 EQUIPMENT QUALIFICATIONS l SDAR: CP-87-36 (INTERIM REPORT) Gentlemen: ' On July 15, 1987, we verbally notified your Mr. Ian Barnes of a deficiency involving evaluations which indicate that the vendor equipment qualification reports for equipment furnished in accordance with-several Balance of Plant
. project specifications contain deficiencies and discrepancies. This is an interim' report of a potentially reportable item under the provisions of l 10CFR50.55(e). l Our evaluation of this issue includes a program to compare specific plant !
service conditions with the applicable test documents. The results of this , program are needed to determine the impact of this issue upon the safety of plant operations. We expect to submit our next report on this issue by November 23, 1987. l l Very truly yours, f I W. G. Counsil ' WJH/mlh c - Mr. R. D. Martin, Region IV ,
- l. Resident Inspectors, CPSES (3) !
I i I L ! L l l l 300 North Obve Street LB SI Dallas, Texas 73:01
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