IR 05000445/1985007: Difference between revisions

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{{Adams
{{Adams
| number = ML20213A032
| number = ML20238A048
| issue date = 04/22/1987
| issue date = 08/12/1986
| title = Ack Receipt of 860402,870128,0220 & 0327 Ltrs Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-445/85-07 & 50-446/85-05
| title = Partially Withheld Memo Discussing Concerns Re Insp Repts 50-445/85-07 & 50-446/85-05.Author Directed & Concurred W/ Insp Plan & Approved Insp Findings.Draft Memos,Ltrs to Util, Draft Insp Repts & Notice of Violation Encl
| author name = Johnson E
| author name =  
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name = Counsil W
| addressee name = Johnson E
| addressee affiliation = TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
| addressee affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| docket = 05000445, 05000446
| docket = 05000445, 05000446
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = NUDOCS 8704270351
| document report number = NUDOCS 8708200352
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| package number = ML20237K807
| page count = 12
| document type = INTERNAL OR EXTERNAL MEMORANDUM, MEMORANDUMS-CORRESPONDENCE
| page count = 48
}}
}}


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o In Reply Refer To: NR 22 E Dockets: 50-445/85 07 50-446/85-05 TU Electric ATTN: W. G. Counsil
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  . Executive Vice President 400 N. Olive L.B. 81 Dallas, Texas 75201
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Gentlemen:
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Thank you for your letters of April 2,1986, January 28, 1987 February 20, 1987, and March 27, 1987, in response to our letters dated February 3,1986, and December 30, 1986. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violation. !ic will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be maintaine
  \    UNf7ED STATES NUCLEAR REGULATORY COf.1 MISSION 5 / ' '
REGloN IV
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811 RYAN PLAZA DRIVE. SUITE 1000 ARLINGTON, TEXAS M011 AUG 12 886 MENORAHDUM FOR:  Eric H. Johnson, Director Division of Reactor Safety and Projects FROM:
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Chief, Reactor Project Section 8. RP8 SULJECT:
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FOLLOWUP DN INTERVIEW WITH NRC IN'SPECTION PERSONNE  -
ASSIGNED STATION (CPSES)
TO THE COMANCHE PEAK STEAM ELECTRIC GENER
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REFERENCE:  itRC IR N /85-07; 50-446/85-05, . dated February 3,1985 Meeting on February 25, 1986, same subject Memorandum, March 25,1986 E. H. Johnson tog dated Me ran um  H. Johnson to  dated As requested in your March 25, 1986, memorancum, and the later request in you June 5,1986, memorandum this documents the conce'rns related specifically to the F CPSES bruary  tlRC Inspection Report No. 50-445/85-07; 50-446/85-05, dated 3, .198 inspection plan and approved the inspection findings.As the Chief CP relayed to the licensee at the conclusion of the inspection.These findirigs were
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The items of concern regarding the subsequent handling of inspection findings were relayed to you for your consideration during discussions prior to the February 25, 1986, meeting conducted in your office. The February 25, 1986, included certain other Region IV DRSP personnel , including myself and ndin  The items of concern regarding the handling of inspection again discussed in some detai Enclosures findings, my 1 through 7 to this memorandum address, as I understand the concern Enclosures 8 and 9 are draft memoranda requesting consideration comitments to these be provided requirement regarding the ASME Code requirements and applicant I appreciate your consideration regarding the items of concern and your ASME Code questions. support in obtaining the needed consideration and assistance r
        .
eactor Pro ect ,
 
===Enclosures:===
Asstated(9)        ,
Attachment:        )
Draft NRC IR No. 50-445/85-07; 50-446/85-05
 
REGION IV
  , A e      PARKWAY CENTRAL PLAZA BulLDING 811 RYAN PLAZA DRIVE. SulTE 1000
  "'s' e ,' ,, ' . */      ARLINGTON. TEXAS 79011
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In Reply Refer To:         .
Dockets:   50-445/85-07 50-446/85-05 Texas Utilities Electric Company        ,
ATTN: M. D. Spence, President, TUGC0 Skyway Tower l    400 North Olive Street          '
Lock Box 81 Dallas, Texas   75201 Gentlemen:
l    This refers to the inspection conducted under the Resident Inspection Program by Mr. H. S. Phillips and others during the period April 1, 1985, through June 21, 1985, of activities authorized by HRC Construction Permits CPPR-125 and CPPR-126 of the Comanche Peak facility, Units 1 and 2, and to the discussion of our findings with Mr. J. T. Merritt, and other members of your staff at the conclusion of the inspectio Areas examined during the inspection included plant status, action on previous NRC inspection findings, action on applicant identified designWithin    construction these deficiencies (10 CFR Part 50.55(e) reports) and plant tour areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors. These findings are documented in the enclosed inspection repor During this inspection, it was found that certain of your activities were in violation of NRC requirement Consequently, you are required to respond to this violation, in writing, in accordance with the provision of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this lette *~'
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        ' lx No,,,  nry je  rede N l* hN MAbed by Me NE      ~&J wnew r  r e r). y ,e a /6SWs' drc k oes/Wer        e Sai/q ,. , y' n may
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Yf#"      C    e/ pp. 4 ,. pg,./ ,/
l'  *mn  Pecu'c Wesponse fem A
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l Texas Utilities Electric Company  2 Should you have any questions concerning this inspection, we will be pleased to discuss them with yo
 
Sincerely,
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===Enclosures:===
Appendix A - Notice of Violation Appendix 8 - NRC Inspection Report 50-445/85-07 50-446/85-05
 
REGION IV==
NRC Inspection Report: 50-445/85-07  Permit: CPPR-126 50-446/85-05  CPPR-127 Docket: 50-445; 50-446 Applicant: Texas Utilities Electric Company (TUEC)  **
l Skyway Tower 400 North 011"e Street Lock Box 81 Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES)
Units 1 and 2
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Inspection At: Glen Rose, Texas
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Inspection Conducted: April 1,1985, through June 21, 1985
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Inspectors:
H. S. Phillips, Senior Resident Date Reactor Inspector Construction (pars. 1, 2, 3, 8, 9, 10, 11, 15, 16, 17, 18, and 19)
J. E. Cummins, Senior Resident Reactor Date Inspector Construction (April 1 - May 10, 1985)  {
  (pars. 1, 3, and 19)    )


Sincerely, Original Signed byr R.E.HALU E. H. Johnson, Director Division of Reactor Safety and Projects
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TU Electric     ;
Date D. M. Hunnicutt, Section Chief
G. S. Keeley, Manager,
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Date Reactor Projects Branch 2 Approved: $ (pars. 1, 4, 5, 6, 7, and 19)    1 D. M. Hunnicutt, Section Chief, Date f,, Reactor Project Section B    '
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Inspection Summary Inspection Conducted April 1, 1985, thecugh June 21, 1985(Report 50-445/85-07)
Areas Inspected: Routine, announced and unannounced inspections of Unit I which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), review of documentation for site dams, and review of 10 CFR Part 21 and 10 CFR Part 50.55(e) construction deficiency status, The inspection involved 77 inspector-hours onsite by  .
four NRC inspector Results: Within the areas inspected, three violations were identified (failure to promptly correct an identified problem with RTE - Delta Potential Transformer Tiltout Subassemblies, paragraph 3.a.; and commercial non-shrink grout was used to grout the Unit 1 reactor coolant pump and steam generator supports in lieu of Class "E" concrete, paragraph 3.b.; hydrogen recombiners out-of-specification voltage recorded on quality release document, paragraph 3.c).
 
Inspection Summary Inspection Conducted April 1, 1985, through June 21, 1985 (Report 446/85-05)
Areas Inspected: Routine, announced and unannounced inspections of Unit 2 which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), review of documentation for site dams, review of documentation for voids behind the stainless steel cavity liner of reactor building, observation of NDE on liner plates, inspection of concrete batch plant, review of calibration laboratory records for batch plant, review of concrete laboratory testing, inspection of level C and D storage, review of reactor pressure vessel (RPV) and piping records / completed work, and review of 10 CFR Part 21 and 10 CFR Part 50.55(e) construction deficiency status, and review of violation and unresolved items statu The inspection involved 355 inspector-hours onsite by four NRC inspector Results: Within the sixteen areas inspected seven violations were identified (failure to provide objective evidence to show that concrete central and truck  f mixer blades were inspected, paragraph 8; failure to issue a deficiency report  {
on cement scales that were out-of-calibration, paragraph 9c; failure to  j translate design criteria into specifications, procedures, and drawings,  l l
paragraph 12a.; failure to maintain RPV installation tolerances / document v  nonconformance, paragraph 12b.; failure to audit RPV specifications,  j procedures, installation, and as-built records, paragraph 12d.; failure to identify a spool piece, paragraph 14b.; and failure to maintain material  ,
I records, paragraph 14 ?
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DETAILS Persons Contacted Applicant Personnel M. McBay, Unit 2 Reactor Building Manager  .
B. Ward, Gen. Supt., Civil D. Chandler, QA/QC Civil Inspector W. Cromeans, QA/QC, TUGC0 Laboratory / Civil Supervisor
  *#J. Merritt, Assistant Project General Manager
  *#P. Halstead, Construction Site QA Manager
  #C. Welch, QA Supervisor TUGC0 (Construction)
J. Walters, TUGC0 Mechanical Engineer K. Norman, TUGC0 Mechanical Engineer J. Hite, B&R Materials Engineer G. Purdy, B&R CPSES QA Manager
  * Denotes those present at May 10, 1985 exit intervie # Denotes those present at June 10, 1985 exit intervie The NRC inspectors also interviewed other applicant employees during this inspection perio . Plant Status Unit 1 At the time of this inspection, construction of Unit I was 99 percent complete. The fuel loading date for Unit 1 is pending the results of f ongoing NRC review l
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Unit 2 At the time of this inspection, construction of Unit 2 was approximately 74 percent complete. Fuel loading is scheduled for approximately 18 months after Unit 1 fuel loadin t I Applicant Action on Previous NRC Inspection Findings
        ! (Closed) Unresolved Item 445/8440-02: Potential Problem with Potential Transformer Tiltout Subassemblie By letter dated June 15, 1983, Transamerica Delaval notified the l applicant of an RTE - Delta 10 CFR Part 21 report to the NRC reporting a potential problem with the primary disconnect clips of the potential transformer tiltout assembly used in the emergency diesel generator control panels at CPSES. The Transamerica Delaval l
letter also provided instructions for correcting the proble i However, the NRC inspector could not determine if the problem had l
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been corrected at CPSES and made this an unresolved item. The applicant determined that the problem had not been corrected and subsequently performed the recommended corrective action. The Unit I corrective action work activities were documented on startup work permits Z-2912 (train A) and 2-2914 (train B). The Unit 2 work activities are being tracked as master data base (MDB) item 3003-3 The failure to promptly correct this identified problem is an apparent violation (445/8507-01; 446/8505-01).
 
. (Closed) Unresolved Item 445/8416-03: Commercial Grout Used in Lieu of Class "E" Concrete The applicant determined that the use of non shrink commercial grout in lieu of the Class "E" concrete specified on drawing 2323-S1-0550 was acceptabl Design Change Authorization 21179 was issued to drcwing 2323-51-0550 accepting the use of the commercial non-shrink grou However, the failure to grout with Class "E" concrete es specified on the drawing at the time the work was accomplished is an apparent violation (445/8507-02). (Closed) Unresolved Item 445/8416-04: Hydrogen Recombiners -
Out-of-Specification Voltage Recorded on Westinghouse Quality Release Document Quality Release N-41424 was revised changing the specified voltage from 10+-2V to 12+-2V which put the questionable voltage within specification limit However, the failure of receipt inspection to verify that the QRN-41424 was filled out accurately as required by Procedure QI-QAP7.2-8 is an apparent violation (445/8505-03). (0 pen) Unresolved Item 445/8432-06; 446/8411-06; Lobbin Report Described Site Surveillance Program Weaknesses During this reporting period the NRC inspector reviewed the status of this open item several times and interviewed TUEC management and site surveillance personnel. The Lobbin report stated that the scope and objectives of the site surveillance program were unclear, lacking both purpose and directio There is no specific regulatory requirement to have a surveillance program; however, TUEC committed to have a surveillance program and has established procedures to implement such a program as a part of the 10 CFR Part 50, Appendix B, QA program. This extra effort is a strength; however, the NRC inspector also observed, as did the Lobbin Report, that the surveillance program lacks both purpose and direction to be effective and complimentary to the audit and inspection program Since the TUEC audit group is not located on site, the TUEC surveil-lance program on site takes on added significanc This item was discussed with the TUEC site QC manager who described 1 a reorganized site surveillance function and changes that have occurred. New procedures which describe this organization's duties and responsibilities are forthcoming.
 
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    -5-TUEC has elected to defer responding to the violations pertaining t the audit function in NRC Inspection Report 445/84-32; 446/84-11, but rather to have the Comanche. Peak Response Team'(CPRT) respond.to this repor^t and other QA matter The surveillance issue is closely tied to the audit deficiencies in NRC Inspection Report No. 445/84-32; 50.446 84-11. This. item will remain open pending the review and implementation of the CPRT action pla A special point of interest , ,
will be how audits and surveillance work together to evaluate the control of all safety related activities on site to essure quality especially the overview of quality control effectivenes . Document Inspection of Site Dams The NRC inspector reviewed documents describing the inspection activities performed on the Squaw Creek Dam (SCO) and the safe shutdown impoundment (SSI) for impounding cooling water for the two units at CPSES. The  !
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ATTN:
purpose of the SCO is to impound a cooling lake for CPSES. A secondary reservoir-(SSI) is formed by a channel connecting the SCO impoundment to the SS Three documented inspections have been performed since 198 The inspections were: Relevant data for'SCD is contained in Phase I Inspection, National Dam Safety Program, Squaw Creek Dam, Somervell County, Texas, Brazos River Basin, inspection by Texas Department of Water Resource Date of Inspection: June 10, 198 Inspection on August 25, 1982, by registered professional engineers from Mason-Johnston & Associates, Inc., and Freese & Nichols, In I Inspection on September 19, 1984, by a registered professional engineer from Mason-Johnston & Associates, In The inspection activities consisted of visual inspections by inspection teams that included accompanying Texas Utilities Service, Inc. (TUSI),
Nuclear Licensing    .i Skyway Tower 400 North Olive Street    l Lock Box 81 Dallas, Texas 75201 Juanita Ellis President, CASE 1426-South Polk Street Dallas, Texas 75224
and Texas Utilities Generating Company (TUGCO) representative Photographs were taken as a part of the documentation. The data for the
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  . piezometer observations and the data for the surface reference monuments were reviewed by applicant personnel and Mason-Johnston engineer No items of significance were observed or reported by these inspection teams. Slight erosion areas were observed and reporte A cracked area on the service spillway upstream right bridge seat was observed by the inspection teams and continued monitoring of this area was recommended by Mason-Johnston and Associate No signs of cracks, settlements, or horizontal movement at any location within the SCO or the SSI were l reported.
    /DRSP RIV/RRI t 21//7 SPhillips:lc RIV/CPTg IBarnes //EHJohnson 4/M/87  &/tt/87 f h//87 8704270351 870422 -
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PDR ADOCK 05000445 0 PDR


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TU Electric  -2-Renea Hicks Assistant Attorney General Environmental Protection Division P.O. Box 12548 Austin, Texas 76711 - 2548 Administrative Judge Peter Bloch U.S. Nuclear Regulatory Commission
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The NRC inspector reviewed the applicant's records and the Mason-Johnston inspection reports. These documents indicated that the SCD and SSI were structurally stable and that the applicant was performing inspection activities to maintain the structural integrity of these dam The state of Texas requires periodic inspections of these dams (principally the SCO) due to inhabited dwellings downstream. The applicant has met these inspection requirement . .
No violations or deviations were identifie . Voids Behind the Stainless Steel Cavity Liner in Unit 2 Reactor Building The NRC inspector reviewed applicant records, including NCR C-82-01202; NCR C-1784, Rev. 1; NCR C-1784, Rev. 2; NCR C-1766, Rev. 1; NCR C 1791, Rev. 1; NCR C-1824, Rev. 1; NCR C-1824, Rev. 2; Significant Deficiency Analysis Report (SDAR) - 26, dated December 12, 1979; DCA-20856; and Gibbs and Hill Specification 2323-55-18. The review of these records and documentation and discussions with various applicant personnel indicated the following:
Structural concrete was placed in Unit 2 reactor building at elevation 819 feet 6-3/4 inches to 846 feet 6 inches on June 21, 1979. This concrete was placed adjacent to the stainless steel liner walls. The concrete forms for this pour were not removed until October 1979 due to subsequent concrete placements for the walls to elevation 860 feet 0 inches. When the forms were removed, honeycombs and voids were observed by applicant personne The applicant's review of the extent of unconsolidated concrete resulted in the issuance of SDAR-26 on December 12, 1979: Investigations were begun and Meunow and Associates (M&A) of Charlotte, North Carolina, were contracted to perform nondestructive testing on in place concrete. M&A performed these tests on a two foot grid pattern on the compartment and liner sides of all four steam generator (SG) compartment walls. The selected test locations did not include the locations where the voids were later found to be located; therefore, the voids were not detected during the M&A testin In August 1982, preparations were made to pour the concrete annulus around the reactor vessel. When the expanded metal formwork was removed from the reactor side of the compartment walls, voids were observed and NCR C-82-01202 was prepare DCA 20856 was prepared as a procedure to repair the void area. DCA 20056 indicated that the voids were not extensive (a surface area of about 28 square feet by  i 8 inches maximum depth) and that the repair procedure assured that the total extent of voids had been identified. One half (0.5) of a  {
I cubic yard of concrete was used to complete the repairs as indicated on grout pour card 26 The applicant's review and evaluation of the gird pattern and a comparison of SG compartments 2 and 3 to 1 and 4 indicated that voids did not exist in SG compartments 2 and 3. The review of test girds extended down to elevation 834 feet, which is the floor elevation of the line The liner walls of SG compartments 1 and 4
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Washington, D. C. 20555 Elizabeth B. Johnson Administrative Judge Oak Ridge National Laboratory P. O. Box X, Building 3500 Oak Ridge, Tennessee 37830 Dr. Kenneth A. McCollom 1107 West Knapp Stillwater, Oklahoma 74075 Dr. Walter H. Jordan 881 Outer Drive Oak Ridge, Tennessee 37830 Anthony Roisman, Es Executive Director Trial Lawyers for Public Justice 2000 P. Street, N.W., Suite 611 Washington, D. Texas Radiation Control Program Director bec: to DMB(IE01)
bec distrib by RIV:
RPB  MIS System RRI-0PS  RSTS Operator RRI-CONST  R&SPB DRSP  R. Martin, RA RIV File  J. Taylor, IE J. Konklin, IE  L. Chandler, OGC D. Weiss,LFMB(AR-2015)  RSB I. Barnes, CPTG
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Log # TXX-6211 File # 10130
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  -7-were not tested at elevation 834 feet, but at elevation 836 feet  !
which is above the area of the identified voids. No testing was  j done on the liner side of the area of the ;ide below elevation 836
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feet. The program also included removal of c 7ch x 2 inch plugs  f I
from the stainless steel liner at locations where test indications raised questions concerning the concrete. The inspections of the concrete by applicant personnel after the plugs were removed confirmed that there were no additional unconsolidated concrete -
areas (voids).
 
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The applicant removed stainless steel liner plates from three areas (one arec about 1 foot by 1 1/2 feet and two areas about 3 feet by 1 foot, excavated or chipped to sound concrete, and cleaned the concrete surface area. One and one quarter inch (11/4) diameter probe holes and grout access holes were drilled in the liner plates to determine the extent of and to assure full definition of the void are Air access holes were drilled in the stainless steel liner plates to assure that grouting would be accomplished in accordance with the procedur The procedure (DCA-20856) specifed that the grout was to be cured for 28 days or until the grout reached a compressive strength of 4000 psi. Repairs to the liner plates were specified in DCA-20856 and G&H Procedure 2323-55-1 DCA-20856 required that under no circumstances was cutting of the liner across weld seams, across embedded weld plates, or into leak chase seal welds or drilling through the liner at leak chase channels, embeds, or weld seams permitte Documentation review indicated that DCA-20856 was adhered to and that no cutting or drilling occurred in prohibited location No violations or deviations were identifie .
6. Nondestructive Testing Observations of Liner Plates in Fuel Transfer Canal The NRC ir;pector observed portions of non-Q liquid penetrant examinations (PT) being performed on liner plate welds following re-installation of the liner plates in the areas of the fuel transfer canal removed for inspection and repair of the concrete. The inspector performed the PT on the welds as required by the repair package and the procedure (QI-QP-11.18-1,
  " Liquid Penetrant Examination"),. Scattered weld porosity was identified by the inspection. The porosity was ground out and a repeat PT was performed. The final inspection is scheduled to be performed by QC inspection personnel. The liner plate areas to be inspected by PT were identified in DCA 2085 No violations or deviations were identifie I
 
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      -8-7. Rebar Placement and Cadweld Splice Observations and Records The NRC inpector reviewed the rebar placement and Cadwell Splice activities associated with the Unit 2 containment (reactor butiding)
closur Calibration of Tensile Tester The NRC inspector observed the' calibration of the Tinus-Olson ,
Universal Testing Machine (Model Number 600-12 Identification Number M&TE-784)'on April 2 and May 7, 1985. The machine was calibrated just prior to performing tensile testing of cadweld splices and
'    subsequent to completion of' tensile testing each day that tensile testing was performed. The machine calibration date for April 2, 1985, prior to start of tensile testing was observed by the NRC inspector and recorded as follows:
Nominal load Calibration Reading Error Error Remarks (lbs)  (1bs)  (lbs) %
0  0  0 0 0 machine on 4/2/85 100,000 99,750  +250 +0.25 200,000 199,600  +400 + ,00 299,450  +550 +0.18 350,000 350,300  -300 -0.08 400,000 401,200  -1200 -0.03 500,000 501,350  -1350 -0.27 600,000 602,450  -2450 -0.40 The NRC inspector reviewed calibration data for March 4, March 8, April 2, April 3, April 30, and May 7, 1985. All calibration data met within the +/- 1% accuracy requirement specified by Calibration Procedura 35-1195-IEI-37, Revision 3, dated March 11, 1982. The reference standards were identified as follows:
ID N Manufacturer Calibration Due Date RS-75 BLH Electronics January 27, 1987 RS-7 BLH Electronics January 27, 1987 v Observation of Cadweld Splice Tensile Testing (1) Qualification Tensile Testing On April 2, 1985, the NRC inspector observed the following tensile testing of cadweld splices for cadwelder qualification:
EBD Q8, GBH Q1, GBH Q2, GBV Q1, BFD Q4, BFD Q3, BFH Q4, GAH Q1, GAV Q1, and GBV Q Each of the above qualification cadweld splices was tensile tested to 400,000 pounds (100,000 psi) and met the requirements stated in the procedur _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
 
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  (2) Production Tensile Testing        ,
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The NRC inspector observed the tensile tester calibrations and      /,-i'
the following production cadweld splices tensile. testing on      i May 7,1985: FXD 3P, FYD 4P, FYD 8P, FRD 87P,.and FUD 6 l
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            , .e Each of the above production cadweld splices was tested to 400,000 pounds (100,000 psi)and met the requirements stated in
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the procedur * '
  (3) Installation of Production Cadweld Splices        I The NRC inspector observed installation of rebar and cadweld      ,
splices at frequent ir,tervals (five or more observations per      ,/
week during the weeks of April 8 and 15; May 6, 13, 20, and 27; and June 3, 1985). The rebar installation for the Unit s closure was performed in the area identified as elevation' 80 ,
feet to elevation 875 feet and azimuth 300 degrees to 335 :      .
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degrees. The installation activities observed included'rebar spacing, location of cadwelds, observation of selection and      \
removal for testing of cadweld splices for testing, and determination of location of rebars and cadwelds for the as-built drawing (4) Documentation Reviewed        e The NRC inspector reviewed the following documentation for the rebar placement and cadwelding for the Unit 2 containment (reactor building) closure area:
Drawings  DCAs NCRS 2323-5-0785, Re , Rev. 1 C85-200294 2323-5-0786, Re C85-200339 Re ,
2323-S1-500, Re C85-200355, Re , Re , Re (Sheets 1-7)      i 2323-52-508, Re ?2 2323-52-506, Re l
 
l No violations or deviations were identifie . Concrete 8atch Plant Inspection l
The NRC inspecor used a nationally recognized checklist to inspect the        !
concrete production facilities This list included the specific        '
characteristics for the foll ning areas: (1) material storage and handling of cement, aggregate, water and admixture, (2) batching equipment scales, weighing systems, admixture dispenser, and recorders, (3) central mixer (not applicable because it had been dismanteled), (4)
ticketing system, and (5) delivery syste l
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TUELECTRIC   b$ - YY
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w ,,,, c. %,   January 28, 1987 t w w,,e ra.e n ,sca,,     FEB - 31987 ! ,
y   ?The current batching is a manual operation since almost all concrete has
U. S. Nuclear Regulatory Commission    "-
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ATTN: Document Control Desk Washington, DC 20555 SUBJECT:
  -been placed. The central mixer was dismanteled and removed from site two
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
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DOCKET NOS. 50-445 AND 50-446 INSPECTION REPORT NOS.: 50-445/85-07 AND 50-446/85-05 SUPPLEMENTAL REQUEST FOR INFORMATION TO NRC NOTICE OF VIOLATION (NOV) ITEMS 1, 2.A AND REF:  TUGC0 LETTER TXX-4727 FROM W.G. COUNSIL TO E. H. JOHNSON DATED APRIL 2, 198 ,
  , or three years ago when concrete placement was virtually complete ~'
Gentlemen:
Presently, the backup batch plant (which was a backup system for the
We have reviewed your letter dated December 30, 1986, requesting additional information to our response referenced above regarding NOV 445/8507-01 and 446/8505-01 (Item 1), 445/8507-02 (Item 2.a) and 445/8507-03 (Item 2.b). We hereby respond to the request for supplemental information in the attachment to this lette We requested and received an extension until January 26, 1987, in providing our response during a telephone conversation with Mr. I. Barnes on January 15, 1987. We requested and received an extension until February 16, 1987, in providing our response to NOV 445/8507-03 (Item 2.b) during a telephone conversation with Mr. I. Barnes on January 23, 1987. We requested and received an extension until January 28, 1987, in providing our response during a telephone conversation with Mr. I. Barnes on January 26, 198 Very truly yours,
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central mixer) is in operation to complete the remaining concrete
W. G. Cougsil
/~  plac qents. This batch plant is in good condition and complied with the f '
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By:- -
subject dmcklist except for one area The NRC inspector inspected the inside of one of three trucks used for . .
G. S. Keeley Manager, Nuclears .evsing vu/
mixing; concrete (that is, the batch plant dispenses the correct weight of materials as required by the specific design mix numbers and the truck then mixes the batch to be placed.) The blades inside the truck are
RSB:lw      V Attachment c - Mr. Eric H. Johnson, Region IV Mr. D. L. Kelley, RI-Regicn IV Mr. H. S. Phillips, RI-Region IV n m/ n r/I I J 'Jf #
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subject to wear and should be checked at a reasonable frequency. The Brown & Rost (B&R) representative responsible for checking the blades in accordance witn B&R Procedure 35-1195-CCP-10, Revision 5, dated December 4, 1978, was asked for evidence that the blades had been checked for wear on a quarterly basis and it was found that there was no record of i  such checks dating back to 1977 when they were initially checke Procedure CCP-10, paragraph 3.10 " Truck Mixing", is silent on blade wear but Section 3.11 infers that the blades should be checked for both central and truck mixing. The inspection of both central and truck mixing blades was not documented, although the B&R representative stated that the mixing blades were periodically inspected and laboratory testing would have probably indicated if there was a problem with the mixing blade Strength and uniformity tests have consistently been within the acceptable range indicating that concrete production was acceptable even though mixing blade inspection was not documente Otherwise, the condition of the inside of the truck was satisfactory as the drum and charging / discharging were clea The water gage and drum counter were in good conditio This failure to follow procedures is a violation of 10 CFR 50, Appendix B, Criterion Subsequent to the identification of this violation, the blades were checked for wear and blade wear was presently within allowable limits (445/8507-04; 446/8505-02).
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No other violations or deviations were identifie Calibration Laboratory for Batch Plant    l The NRC inspector obtained batch plant scale numbers from tags which  l indicated that the scales had been calibrated and were within the calibration frequenc Cement (MTE 779), Water (MTE 766), admixture scale (MTE 764), and aggregate (MTE 780) were reviewed. The scales had been periodically calibrated since the batch plant was activate The records were adequate except as follows:   ! Scales MTE 766 records do not clearly differentiate between the required accuracy of the scale and the digital readou *
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    -11- Scales same MTE 779 and 780 records show various accuracy ranges for the scale; i.e., MTE 779 (SN749687) records the following: report dated January 1976 gives EE; report dated July 1976 gives 1% while report October 1976 gives +/- 0.2%.
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The above items are unresolved pending further review of the licensee actions regarding these records during a subsequent inspection (445/8507-05; 446/8505-03). The calibration appeared to be prope *
      * * Records July for scales 16, 1975, MTE 779 records contained B&R memo IM-1108 d which described a nonconforming conditio This condition affected the water and cement scales causing a 24-48 pound deviation during the calibration tes The memo stated that the condition was corrected and the scales were then calibrated; however, no deficiency report was written as required by B&R Procedure CP-QCP-1.3, " Tool and Equipment Calibration and Tool Control" dated July 14, 1975, and CP-QAP-15.1, " Field Control of Nonconforming Items," dated July 14, 1975. As a result there is no evidence if concrete that corrective production action included was adversely affected.an evaluation to determine This failure to assure that a nonconforming condition was evaluated is a violation of Criterion XV of 10 CFR Part 50, Appendix (445/8507-06; 446/8505-04).
 
1 Concrete Laboratory Testina TUGC0 procedure QI-QP-11.1-1, Revision 6, was compared with ASME Section III, Division 2, Subsections 5222, 5223 and 5224 to assure that each ASTM testing requirement was incorporated into the procedur The NRC inspector inspected the testing laboratory equipment and found the test area was in good condition and each piece of equipment was tagged with a calibration sticker which showed it to be within the required calibration frequenc test requirements and equipment. Test personnel were knowledgeable of Thefollows:
as NRC inspector witnessed field tests performed by laboratory personnel Date Truck No. Mix No. Ticket N Air Content (%) Slump Temperature (in.) _F 6/3/85 RT-41 925 64013 Req 8.2-1 NA 70 max Mea 8.7- NA 57 6/3/85 RT-35 128 64014 Req 5.0- max 70 max  <
Mea .25* 57
* Truck was rejected by quality control but was later accepted when second slump reading came into required rang E
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The following laboratory equipment was checked and found to be within calibration: Forney Compression Tester, MTE 3031; Temperature Recorder ;
MTE 3013 and 3014; Unit Volume Scale, MTE 1053; Pressure Meters MTE i 3000B, 3002 and 3004; Sieves MTE 1286, 1239, 1272, 1274, 1136A, 1156, 1094, 1093, 1095, 1178, 1179, 1300 and 1180; Aggregate scales, MET 1058 and 1067; and 2" grout mold MTE 111 The following test records for placement number 201-5805-034 were reviewed: (1) concrete placement inspection, (2) concrete placement -
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summary and, (3) unit weight of fresh concret l No violations or deviations were identifie . Inspection of Level C and D Storage I
The NRC inspector inspected all laydown areas where ciping, electrical conduit, cable, and structural reinforcing steel were store These materials were neatly stored outside on cribbing in well drained areas which allowed air circulation and avoided trapping wate This met the Level "D" storage requirements of ANSI N45. The electrical warehouse contained miscellaneous electrical hardwar This building was required to be fire and tear resistant, weathertight, and well ventilated in order to meet Level "C" storage requirements. This warehouse was well kept and met all requirements except for a lock storage area located upstairs at the rear of this building (electrical termination tool room). Two minor problems were identified and the warehouse personnel initiated action to correct the The first problem noted was that a box of nuclear grade cement was marked
  " shelf life out of date" but it had no hold ta The box was subsequently tagged with Nonconformance Report (NCR) E85-200453 af ter being identified by the NRC. During discussions with the warehouseman, the NRC determined that engineering told the warehouseman to mark the material and lock it up, but did not tell him to apply an NCR or hold ta TUEC should determine if engineering is aware of nonconforming material controls and provide training if this is other than an isolated instance. Also, the NRC inspector noted a very small leak in the roof above the electrical termination tool room. This lesk was in an area that did not expose hardware to moisture. The roof is currently being repaire The millwright warehouse storage area was inspected; however, only a small number of items or materials were stored in this area. The overall storage conditions in this area met or exceeded Level "C" storage requirement No violations or deviations were identifie . Reactor Pressure Vessel and Internals Installation - Unit 2 This inspection was performed by an NRC inspector to verify final placement of the reactor pressure vessel (RPV) and usernals by examining the completed installation and inspection record I
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400 North Othe Street I..B. 81 Dallas Te as 75201
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a. Requirements for Placement of RPV Requirements for placement of the RPV to ensure proper fit-up of all other major NSSS equipment are in Westinghouse Nuclear Services  ,
Division (WNSD) " Procedure for Setting of Major NSSS Components", )
Revision 2, dated February 13, 1979, and " General Reactor Vessel Setting Procedure" Revision 2, dated August 30, 1974. The NRC  '
inspector reviewed the following drawings, which were referenced in the RPV operation traveler, to verify implementation of WNSD recommendation . j o WNSD drawing 1210E59 " Standard - Loop Plant RV Support Hardware Details and A,sembly" o WNSD drawing 1457F27 " Comanche Peak SES RCS Equipment Supports
  - Reactor Vessel Supports"    .
l o CE drawing 10773-171-004 " General Arrangement Elevation"  l o CE drawing 10773-171-005 " General Arrangement Plan" l
Neither site prepared installation drawings nor specifications (which implemented the WNSD recommended procedures) were available and the drawings examined did not show certain specific installation criterion such as centering tolerances, levelness tolerances and clearance between support brackets and support shoe The lack of engineering documentation did not provide full control of the action and would allow changes to installation criteria important to safety to be made without complying with established change procedure This is considered a violation of 10 CFR 50, Appendix B, Criterion III (446/8505-06).


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b. Document Review The NRC inspector reviewed B&R Construction and Operation Traveler ;
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No. ME79-248-5500 which described the field instructions for  ]
Attachment to TXX-6211
installation of the Unit 2 RP Requirements recommended by WNSD 1 procedures were implemented in the traveler. Worksheets attached to the traveler showed the RPV to be centered and leveled within the established tolerances, Traveler operation 19 required verification of a 0.020 to 0.005 inch clearance between the support bracket and )
  .' January 28, 1987
support shoe, after applying the shim plates. Change 5 subsequently ]
. Page 1 of 4
changed the clearance to a 0.015 to .025 inch clearance. The installation data reflected in attachment 3B of the traveler indicated an as-built clearance of 0.012 to 0.026 inch which exceeds both the original and revised tolerances. This condition was accepted on the traveler based on Westinghouse concurrence, and there were neither nonconformance reports nor documented engineering evaluations to determine if the condition was acceptable. This failure to document nonconforming conditions and engineering deviations is a violation of 10 CFR 50, Appendix B, Criterion XV (446/8505-07)
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The NRC inspector reviewed the following receiving records for RPV hardware and found them to be in order:
  . NOTICE OF VIOLATION ITEM 1 (445/8507-01 AND 446/8505-01)
SUPPLEMENTAL REQUEST FOR INFORMATION We find that the corrective steps taken and results achieved do not indicate that the installation of the replacement RTE-Delta Potential Transformer tiltout assembly was reinspected and accepted. Also, there is no indication that these deficiencies were documented on a nonconformance report for Units 1 and SUPPLEMENTAL RESPONSE TO ITEM 1 In our previous response we noted the deficiencies involving the Unit 1 RTE-Delta Potential Transformer Tiltout assemblies were corrected using startup work permits. We failed to note that inspections and the acceptance of completed work are inherent in the startup work permit program. We have ~
confirmed these activities were accomplished as documented on the completed work permits (Z-2912 & Z-2914).


As noted in our review of a previous NRC finding (445/8407-01), the program for managing 10CFR21 defects was not positively controlled. In some cases, and in the case of the RTE-Delta 10CFR21 notice involved in this finding, the deficiencies were corrected without nonconformance report Under our current program, the Site Coordinator, whose responsibilities are specified in Procedure NE0-CS-1, " Evaluation of and Reporting of Items / Events Under 10CFR21 and 10CFR50.55(e)," is responsible for assuring proper deficiency documentation, including NCRs, are issue Our previous response incorrectly indicated that the Unit 2 10CFR21 deficiencies were corrected in August 1985. The work was actually completed in September 1985 and QA/QC review of the work packages was completed in April 198 l l
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Attachment to TXX-6211
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January 28, 1987 Page 2 of 4


NOTICE OF VIOLATION    i ITEM 2.a (445/8507-02)   !
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SUPPLEMENTAL REQUEST FOR INFORMATION  l The described corrective steps taken do not indicate whether a nonconformance report was written for Units 1 and Follow up by the NRC inspector identified that the drawing which was used for the activity (i.e.,
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Westinghouse Drawing 1457F29) had been reviewed and stamped by Gibbs and Hill, signifying apparent approval of a drawing which allowed use of grout instead of Class E concrete. Accordingly, it is requested that the stated reason for
  .'o Report No.14322 for 54 each closure studs, closure nuts, and closure washers o Report No. 09507 for vessel S/N 11713, Closure Head 11713 an Rings o Deviation notices and corrective action statements The NRC inspector reviewed the following completed travelers'for   -
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internals installation and found them to be satisfactory; o ME-84-4641-5500, " Assemble Upper Internals" o' ME-84-4503-4000, " Install and Adjust Roto Locks o ME-81-2145-5500, "Retorque UI Column Extension" o RI-80-385-5500, " Transport and Install Lower Internals" o ME-84-4617-5500, " Repair Lower Internals" o ME-84-4640-5500, " Assemble Lower Internals" Visual Inspection At this time, visual inspection of the internals by the NRC inspector was not possible, and inspection was limited on the vessel placement to a walk-around beneath the vessel to inspect the azimuth markings and for construction debris between the vessel and cavity. No problems were identified in this are Records of QA Audits or Surveillance
the violation be reevaluated, in that the engineering interface or review may have contributed to causing this violation. Please also identify the design change authorization, if any, that permits use of grout rather than Class E concrete in Unit 2, in that the referenced design change DCA-21,179 would appear to be applicable to Unit I only. In your corrective steps to avoid recurrence, you stated that operational travelers issued during the same time frame for similar type installations would be reviewed for similar deficiencies. Please clarify how this action is pertinent to precluding-recurrence. Your response also indicated that any corrective actions deemed necessary, as a result of this review, would be reported to Region IV by April 15, 1986. In that no supplementary response has been received by Region IV, please identify whether or not your review determined corrective actions  -
were necessar SUPPLEMENTAL RESPONSE TO ITEM As a result of discussions with the NRC Resident Inspector, and our review of the Violation, we have determined that a revision to our response is require The revised response below provides the supplemental information requeste . Reason for Violation We admit to the Violation, ano the required information follow This violation was the result of inadequate design review of the
. Westinghouse equipment drawing by the primary on-site design group after~
the drawing had been received on-site. The drawing, which specified the
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use of grout, disagreed with the Gibbs and Hill, Inc. civil drawing which
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specified Class "E" concrete. Subsequently, this disparity was overlooked when an operational traveler was developed which specified the use of grout without formal design concurrence on incorporatio !
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__g m- 4 1__ <-  -- b+4 m- = ,
The NRC inspector requested TUGC0 QA audits or surveillance performed by TUGC0 of the Unit 2 RPV installation. TUGC0 did not make available any audit or surveillance reports of specifications for placement criteria, placement procedures, hardware placement, or  i as-built records. Failure to perform audits or surveillance of RPV specifications, procedures and installation is a violation of    i 10 CFR 50, Appendix B, Criterion XVIII (446/8505-08).
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No deviations were identified; however, three violations were identified and are described in the above paragraph . Reactor Vessel Disorientation On February 20, 1979, the applicant reported to the NRC Resident Inspector that a design error had resulted in the reactor support structures being placed in the wrong position on the reactor support pedestal such that the reactor would be out of position by 45 degree Initially, Unit 2 was to be a mirror image of Unit 1, however, a design change was initiated to permit identical components for both units. The I
Attachment to TXX-6211          l
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January 28, 1987          !
Page 3 of 4          )
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      - SUPPLEMENTAL RESPONSE TO ITEM 2.a CONT'D Corrective-Action Taken After reviewing grout card No. 186, which documented the material used in the actual installation, and the results of.the compressive strength test


for the grout used in this application, a design change (DCA-21,179) was
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used to document acceptance of the installation. These actions were
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,   initiated after identification of the condition via an NRC Inspection i    Report item.
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design change was implemented for the reactor vessel, but not for the pedestal support location The problem was not considered by the applicant to be reportable under provisions of 10 CFR Part 50.55(e) since the error could not have gone undetecte The problem was reported to the NRC Office of Inspection and Enforcement on February 22, 1979, and during a March 27, 1979, meeting in Bethesda, Maryland, the applicant presented the proposed redesign and rework ,
procedures for relocating the pedestal support No unresolved safety concerns with the repair were identified at the meetin During this inspection the NRC inspector reviewed various documentation relative to the disorientation problem, including design changes and the
}  construction traveler which implemented the repai The following documents were reviewed:
o NRC Inspection Report 50-446/79-03 o NRC Inspection Report 50-446/79-07 o NRC Inspection Report 50-446/79-13  -
o TUSI Conference Memo, dated March 1, 1979, H. C. Schmidt to S. Burwell (NRC Licensing PM)
o TUGC0 letter TXX-2980, dated April 30, 1979, to W. C. Seidle o NRC letter to TUGC0 dated May 29, 1979 o DCA 3872, Revision 1, dated February 28, 1979, Subject: Rework of Structure for Placement of the RPV Support Shoes I
o DCA 4122, dated March 22, 1979, Subject: Replacement of Rebar for RPV Supports o Construction Traveler CE79-018-5505, dated March 14, 1979, Subject:
Rework of Reactor No. 2 Cavity - New RPV Support Locations o Grout Replacement Cards No. 007,008, 009, 010, 014 and 015, various dates, Subject: Replacement of Grout around Rebar for Repair of RPV
  ,, Support Shoes o Various Inspection Reports for Grout Properties and Application for RPV Support Shoes No violations or deviations were identifie . Reactor Coolant Pressure Boundary (RCPB) Systems The inspection was performed to verify: the applicants system for preparing, reviewing, and maintaining records for the RCP8 piping and components; that selected records reflected compliance with NRC requirements and SAR commitments for manufacture, test and installation


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After further review of this matter,~.two DRs (C-86-12 and C-86-13) were issued December 29, 1986, to ensure appropriate drawing changes were made;
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    -16-of items; and that as-built hardware was adequately marked and traceable to records. The following items were randomly selected and inspected:
l Pressurizer Safety Valve - This item was inspected to the commitment-stated in FSAR, Table 5.2-1 which includes ASME Section III, 1971  I edition through winter 1972 addenda. Valve S/N N56964-00-007, which is installed in the B position was inspected. The following records  j were reviewed:
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o QA Receiving Inspection Report No. 21211  <
o Code Data Report Form NV-1 o Valve Body CMTR The valve was in place, however, installation had not been completed; therefore, the hardware installation inspection consisted of verifying that the item was traceable to the record CVCS Spool Piece 3Q1 - Requirements for this item are stated in ASME, .Section III,1974 edition through summer 1974 addenda, which is the commitment from the FSAR, Table 5.2-1. The item was field fabricated from bulk material and installed in the CVCS with field welds number 1 and 3 (ref. BRP-CS-2-RB-076). The following records were reviewed:
o B&R Code Data Report o Field Weld Data Card o NDE Reports o QA Receiving Reports (for bulk order)
o Certified Material Test Report (CMTR)
The installed spool piece was inspected for weld quality and to verify that marking and traceability requirements had been me The item had been marked with the spool piece number (3Q1) and the B&R drawing number, however, marking of the material specification number and type, heat code, or other means of traceability could not be found. In respect to material requiring a CMTR, (nominal pipe size greater than 3/4 inch) NA-3766 requires marking with the applicable specification and grade of material and heat number or heat cod When material is divided, the identification marking is required to be transferred to all pieces. This failure to identify material marking is a violation 10 CFR 50, Appendix B, Criterion VIII (446/8505-09). Loop 3 RC Cold Leg - Requirements for this item are stated in ASME,  ,
Section III, 1974 edition through summer 1974 addenda, which is the commitment from the FSAR, Table 5.2- This piping subassembly
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to formally document acceptance of the grout used in these and other
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nearby installations; and to specifically identify the root cause of this
!    deficiency.


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  ' Action to Prevent Recurrence
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!    Operational travelers, grout cards, and drawings used during the same time
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      -r-17-consists of a 27.5 inch cast pipe with a 22 degree elbow on the reactor end, a 10 inch 45 degree nozzle, a 3 inch nozzle, and three 2 1/2 inch thermowell installation bosses. The following records were reviewed for the subassembly; o QA Receiving Inspection Report No. 12389 o Westinghouse Quality Release (QRN 47523)  -
d "o ' Code Data Report Form NPP-1 o 27 1/2 inch line CMTR o 3 inch nozzle CMTR o Field Weld Data Cards o NDE Reports (1) The NRC inspector determined that CMTR's were required by the code but were not available for the following items:
  . 22 degree elbow
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10 inch 45 degree nozzle
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frame were reviewed for similar installations.- The operational traveler
2 1/2 inch thermowell bosses This failure to maintain retrievable records is a violation of 10 CFR 50, Appendix B, Criterion XVII (446/8505-10).
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review completed on May 15, 1986, indicated that other drawings used for similar installations did not contain the same disparity as discussed in Item 1 above. However, the review revealed that grout was used in seven other installations in Unit I and eight installations in Unit 2 due to
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. differences between drawings 2323-SI-0550, 2323-S2-0550, and 1457F29. The grout used in these other fifteen installations was evaluated and found to l    have a compressive strength in excess of Class "E" concrete. DCA-21,179
,   and DCA-26,101 issued November 8, 1986, and November _26, 1986,
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respectively, were used to document acceptance of the installations in Unit I and Unit 2, respectively. Based on the specific nature of the
. discrepancies and the results of our review, this violation is considered to be an isolated occurrence.


l    A Project Directive was issued on July 17, 1986, reemphasizing, a) the i    requirements for properly documenting design alternatives prior to j    initiating construction activities, and b) the precedence of design versus vendor documents.
(2) Sandusky Foundary and Machine Company test report for the cold leg pipe certifies that material meets requirements of ASME Section II, 1974 editions through winter 1975. Southwest Fabrication and Welding Company code data report NPP-1 Form certified that the cold leg subassembly met requirements of ASME Section III, 1974 edition through winter 1975. The FSAR commitment is ASME Section III, 1974 edition through summer 1974. This discrepancy is unresolved pending the applicant's evaluation to determine if material nonconformances exist (446/8505-11).


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(3) The cold leg NPP-1 Form stated that no hydrostatic (Hydro) test had been performe In discussions with Westinghouse and B&R personnel, the statement was made that it is normal practice to defer the partial hydro test until the whole system is hydro tested. B&R Procedures CP-QAP-12.1 and CP-QAP-12.2 describe requirements for the tes ASME, Section III, (NB-06114(a)) states in respect to the time of testing piping subassemblies, that the component test, when conducted in accordance with the requirements of NB-6221(a)
shall be acceptable as a test for piping subassemblie NB-6221(a)
om m ss __n-mo_ statesm__m that completed components shall be subjoeBod e - - - -- - -


The review of vendor drawings (except for Fire Protection and HVAC
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. drawings) is performed by Stone and Webster Corporation (SWEC) in      '
accordance with SWEC Procedure PP-053, " Review of Vendor Drawings." The review of vendor HVAC drawings is performed by Ebasco in accordance with TUGC0 Procedure ECE-DC-5, " Vendor Document Review." These vendor drawings are reviewed against design drawings to resolve any discrepancie The review of vendor Fire Protection drawings is performed by Impell e Corporation in accordance with Impell Procedure IMP-FP-19, " Vendor
:  ; Document Review." SWEC Procedure PP-053 and Impell Procedure IMP-FP-19 l
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  * will be revised to clarify and specifically require the review of vendor
[  l drawings to insure compliance with design drawing ? Date of Comoliance i
CPSES is currently in full complianc SWEC Procedure PP-053 will be i    revised and. issued by February 27, 198 Impell Procedure IMP-FP-19 will i    be revised and issued by April 30, 1987.


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Paragraph (b) of NB-6221 requires pressure testing of all pressure retaining components that are within the boundary protected by overpressure protection devices and paragraph (c)
permits substitution of the system test for component test. It is not evident that the system test substitution was permitted for pipe subass'embly since NA-1200 makes a distinction in the
  . definition between components and piping subassemblies. NB-6115 states that pressure testing of components shall be performed prior to initial operation of a system and that the Data Repor .
Form shall not be completed nor signed by the code inspector and i the component shall not be stamped until the component manufac-
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turer has conducted the hydrostatic pressure test. Additionally, NA-8231 prohibits the application of a Code stamp prior to hydrostatic test regardless of whether the test was parformed prior to or after installation of the ite The NPP-1 Form for the cold leg had been completed and signed by the manufacturer and Authorized Nuclear Inspector (ANI). An ASME Code Stamp had also been applied to the item. The NRC inspector observed that a hydrostatic test had not been performed and was noted on the NPP-1 for The above items are unresolved pending clarification of code requirements by NRC headquarters (446/8505)-12).


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(4) Since the cold leg pipe subassembly had not been pressure tested prior to installation, the NRC inspector reviewed the procedures and hydro test data applicable to Unit 1, since Unit 2 hydro had not been completed. Requirements for the tests were presented in Procedures CP-QAP-12.2, " Inspection Procedure and Acceptance Criteria for ASME Pressure Testing" and CP-QAP-12.1,
Attachment to TXX-6211
  "ASME Section III Installation, Verification, and N-5 Certifi-cation." Procedure CP-QAP-12.1 requires that a data package to be used in the test, be prepared with the test boundary and the additional following data shown:
. January 28, 1987 Page 4 of 4 o .
o Base metal defects in which filler material has been added, and the depth of the base metal defect exceeds 3/8 inch or 10% of the actual thickness, whichever is less, o Untested vendor performed piping circumferential weld o Approximate location and material identification and description for permanent pressure boundary attachment with applicable support number reference o Weld history, which shall reflect weld removal and/or weld repai _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ .
NOTICE OF VIOLATION ITEM 2.b (445/8507-03)
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SUPPLEMENTAL REQUEST FOR INFORMATION With respect to the corrective steps taken, please describe what formal actions were taken (e.g., specification revision) to justify the change in output voltage from 10 +/- 2v to 12 +/- 2v on the revised Westinghouse Quality Release. Your response also fails to provide any actions taken to assure that other equipment received from this vendor, irrespective of time frame, did not exhibit similar documentation deficiencie SUPPLEMENTAL RESPONSE TO ITEM TU Electric Quality Assurance is in the process of obtaining related information from the vendor and reviewing site documentation. We anticipate submitting our response no later than February 16, 198 '
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The completed hydro data package (PT-5501) for Unit 1, loop 3 .
cold leg was reviewed for compliance with the above requirements. Drawing No. BRP-RC-1-520-001 had been used to annotate the test boundar A handwritten statement on the ,
          '
drawing indicated: "No major base metal repairs could be located" and "No hangers with weld attachments could be located." Welds performed by the pipe subassembly vendor, including the 22 degree circumferential weld and the penetration fittings had not been identified. The following .
items were unresolved regarding the adequacy of the hydro test:
o Was the determination of no major base metal repairs based on a visual inspection or on a review of vendor and site
    '
inspection and repair records?
o Was the shop circumferential weld attaching the 22 degree elbow to the pipe assembly inspected during the test? If so, where is the inspection identified?
o Procedure CP-QAP-12.1 does not require identification of welds for penetrations into the pipe assembly and they were not identified on the drawing. Were those welds inspected? If so, where is the inspection documented?
The above issues will remain unresolved pending further evaluation by the applicant (44S/8507-07; 446/8505-13). Personnel Qualifications - Personnel who had performed selected tasks were identified during inspection of installation records. Training and experience records for the personnel were reviewed to verify that employee qualifications and maintenance of records were current and met requirement Names or codes for five welders and two NDE examiners, who had performed tasks during installation of the items being inspected, were identified and their qualification records  i reviewed. There were no questions in this area of the inspectio . Special Plant Tours (Unit 1 and Unit 2)
On May 23, 1985, the NRC inspector conducted a tour of selected areas of Unit 1 and Unit 2. The group consisted of one NRC inspector, two NRC Technical Review Team (TRT) representatives, two allegers, and several TUEC representative The TUEC representatives tagged each area where a deficiency was alleged. With the alleger's consent, a tape recorder was also used to note locations and describe any alleged deficiencies. The  <
          !
allegers indicated that they had identified all deficiencies during the tour and all other deficiencies that they had knowledge The NRC TRT is l
          >
analyzing this information and will decide whu Action, if any, should be take (
During this tour the NRC inspecter independently identified a  l
          '
questionable practice in that the top of the the pipe chase at the north end of room 88 in Unit 1, safeguards building had two large stickers which


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i-20- stated that areas on the wall were reserved.for pipe hangers GHH-51-1-5B-038-006 and R1(?)1-087-X1 These stickers were dated 198 ];
    =  Log # TXX-6284 P9  File # 10130
It was ??? evident whether hangers missing or none were needed in these locations and the reserve tags were not removed. TUEC representatives were unable to. answer the question immediately. This item is unresolved    3 pending further review during a subsequent inspection. (445/8507-07;    .]
    - -- C  IR 85-07 r C  IR 85-05 7tlELECTRIC  Ref # 10CFR2.201 DwI ,Yjr,,a,,,,    February 20, 1987 U. S. Nuclear Regulatory Commission    3[Q@$Ok l t- i, i; ATTN: Document Control Desk    is '
l 446/8505-05).
I Washington, DC 20555
        [ S 24N L. h SUBJECT: COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)  L- 'h ,
DOCKET NOS, 50-445 AND 50-446    i INSPECTION REPORT NOS.: 50-445/85-07 and 50-446/85-0a SUPPLEMENTAL REQUEST FOR INFORMATION TO NRC NOTICE OF VIOLATION (N0V) ITEMS 1, 2.a and REF: 1) TV Electric Letter TXX-4727 from W. G. Counsil to E. H. Johnson dated April 2, 1986 2) NRC letter from E. H. Johnson to W. G. Counsil dated December 30, 1986 3) TV Electric Letter TXX-6211 from W. G. Counsil to NRC dated January 28, 1987 Gentlemen:
We reviewed your letter (Reference 2) requesting additional information on the subject inspection report, and responded to the NOVs in our letter TXX-6211 (Reference 3). In our response to NOV 445/8507-03 (Item 2.b), we stated that TV Electric Quality Assurance was in the process of obtaining related information from the vendor and reviewing site documentation, and we anticipated submitting our response no later than February 16, 1987. We requested and received an extension in providing our response to NOV 445/8507-03 (Item 2.b) until February 20, 1987, during a telephone conversation with Mr. I. Barnes on February 17, 1987. We hereby respond to Item 2.b in the attachment to this lette .
Very truly yours,
      @
W. G. Counsil RSB:lw Attachment c - Mr. E.'H. Johnson, Region.IV~
Mr. D. L. Kelley, RI - Region IV      l Mr. H. S. Phillips, RI - Region IV Omd1 qw < v7Tl (fh(fY& 00'- 1
  :      ;


        )
No violations or deviations were identifie . ,
400 North Olise Street I..li 81 I)aila3, Texas 75201
1 Routine Plant Tours (Units 1 and 2)
      - - -
At various times during the inspection period NRC inspectors conducted general tours of the reactor building, fuel building, safeguards building, electrical and control building, and the turbine buildin During the tours, the NRC inspector observed housekeeping practices, preventive maintenance on installed equipment, ongoing construction work, and discussed various subjects with personnel engaged in work activitie l I
No violations or deviations were identifie . Review of Part 21 and 10 CFR 50.55(e) Construction Reports Status The NRC inspector reviewed all reports issued to date to assure that NRC and TUEC status logs were complete and up to date. A total of 183 reports have been submitted to date. This inspection period one Part 21 report on Diesel Generator Oil Plugs and two 10 CFR 50.55(e) reports on the Equipment Hatch Cover and SA106 Piping (light wall; were submitte No violations or deviations were identifie ,
1 Review of Violation and Unresolved Item Status The NRC inspector reviewed all violations and unresolved items reported to date to assure that NRC and TUEC status logs were complete and up to In addition, a trend date. Two hundred nineteen items were reviewe analysis of NRC findings was performed to generally determine how many findings could be broadly classified under each criterion of 10 CFR, Part 50, Appendix The frequency of findings showed broad and general trends under the following criteria: II. QA Program; III. Design Control; V. Instructions, Procedures and Drawings; VII. Control of Purchased Material, Equipment and Services; IX. Control of Special Processes; Inspection; XI. Test control; XIII. Handling Storage and Shipping; XVI QA Records; and XVIII Audits. The most significant Also, a trends were number of violations Criterion III, V, VII, IX, X, and XVII occurred with respect to 10 CFR 50.55(e) item These findings mainly pertained to Unit 1 and related closely to trends identified by the NRC Technical Review Team TRT. These trends will be considered during followup on TRT findings. Also, Unit 2 inspection emphasis will consider these trends during future inspection No violations or deviations were identifie _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ _


      . _ _ _ _ _ _ _ _ _
..      ;
I
!        !
.      1-21-    __ _!
19. Exit Interviews The NRC inspectors met with members of the TUEC staff (denoted in paragraph 1) on May 10 and June 10, 1985. The scope and findings of the inspection were discussed. The applicant acknowledged the findings.
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) In Reply Refer To:
Dockets: 50-445/85-07 50-446/85-05 Texas Utilities Electric Company
) ATTN: M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street    '
Lock Box 81 Dallas, TL-as 75201 Gentlemen:
This refers to the inspection conducted under the Resident Inspection Program by Mr. H. S. Phillips and others during the period April 1, 1985, through June 21, 1985, of activities authorized by NRC Construction Permits CPPR-125 and CPPR-126 of the Comanche Peak facility, Units 1 and 2, and to the discussion of our findings with Mr. J. T. Merritt, and other members of your staff at the conclusion of the inspectio Areas examined during the inspection included plant status, action on previous NRC inspection findings, action on applicant identified design construction deficiencies (10 CFR Part 50.55(e) reports) and plant tours. Within these .
k areas, the inspection censisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors. These findings are documented in the enclosed inspection repor During this inspection, it was found that certain of your activities were in violation of NRC requirement Consequently, you are required to respond to this violation, in writing, in accordance with the provision of Section 2.201 l of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal  !
Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this lette I l
RPB2/8 C/RPB2B C/RPB2 NRR HSPhillips/dc DHunnicutt  VNoonan
/ /85 / /85  8 / /85 l
_ _ _ _ - _
e ** e
      .:- . )
Texas Utilities Electric Company 2 Should you have any questions concerning this inspection, we will be pleased to discuss them with yo
Sincerely,
_
_
    .  :
      ..
      ''
eacto Project Branc
    \
Enclosures: Appendix A - Notice of Violation Appendix B - NRC Inspection Report 50-445/85-07 50-446/85-05 cc w/ enclosure:
    *
Tsxas Utilities Electric Company Skyway Tower 400 North Olive Street L ck Box 81 Dallas, Texas 75201    -
      {
Texas Utilities Electric Company ATTN: J. W. Beck, Vice President Skyway Tower 400 North Olive street Ltck Box 81 Dallas, Texas 75201 bec distrib. by RIV:
TEXAS STATE DEPARTMENT OF HEALTH bec: to DMB(IE01)
bec distrib by RIV:
SRPB1  * Resident Inspector OPS
"RPB2  * Resident Inspector CONS  .
R. Martin, RA  * D. Hunnicutt, Chief, RPB2/B 6C. Wisner, D/DRSS PA0 V. Noonan, NRR S. Treby, ELD MIS SYSTEM  RIV File Ju:nita Ellis  Renea Hicks CD. Weiss, LFMB (AR2015)
*w/766 l    .
      )
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Attachment to TXX-6284 February 20, 1987 Page 1 of 2 NOTICE OF VIOLATION ITEM 2.b (445/8507-03)
          .  .
SUPPLEMENTAL REQUEST FOR INFORMATION With respect to the corrective steps taken, please describe what formal actions were taken (e.g., specification revision) to justify the change in output voltage from 10 +/- 2v to 12 +/- 2v on the revised Westinghouse Quality Release. Your response also fails to provide any actions taken to assure that other equipment received from this vendor, irrespective of time frame, did not exhibit similar documentation deficiencie SUPPLEMENTAL RESPONSE TO ITEM Further evaluation of the apparent violation revealed a need to revise our original response. Accordingly, the following revised response is submitte . Reason for the Violation The root cause of the out-of-tolerance condition was a relocation of the Westinghouse subtier supplier's test facility in 1978. A change in the local power level from 240v nominal at the old test facility to 277v nominal at the new location was not initially detected. This change affected the test result output. The supplier representative failed to detect the apparent out-of-tolerance condition noted on the test report prior to release of the equipment from the subtier suppliers shop (Halmar Electronic Corporation) and subsequent shipment to CPSE The out-of-tolerance conditions noted on the test report was not detected by CPSES QC receiving inspection personnel because Westinghouse supplied non-ASME equipment is receipt accepted based on the Westinghouse Quality Release (QR) being complete and general conformance verification to purchase order requirement . Corrective Steos Taken and Results Achieved   j Westinghouse Engineering performed a review of test data for all Electric Hydrogen Recombiner (EHR) Silicon Controlled Rectifier (SCR) panels tested during the period May 13, 1975 to March 8, 1981, bracketing the period during which the 240 to 277 volt plant power transition effect on test results may not have been recognized. The result of the documentation ,
                  -
review for the 209 SCRs revealed an additional out-of-tolerance condition I of the same nature, which has been documented on a Westinghouse Deviation Notice (DN). The Westinghouse Engineering review has established that there was no adverse effect on the performance or function of the equipment as a result of the deviatio ,      !
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      .
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  .- Attachment to TXX-6284 February 20, 1987 Page 2 of 2 NOTICE OF VIOLATION ITEM 2.b (445/8507-03)
                .
SUPPLEMENTAL RESPONSE TO ITEM 2.b - CONT'D Corrective Steos Taken and Results Achieved - Cont'd The output voltage specification was changed and documented in Halmar test procedure TP-001 and approved by Westinghouse Engineering per Westinghouse DN No. 26961 to compensate for the change in the test power suppl . Corrective Steos Which Will Be Taken to Avoid Further Violations Westinghouse's revaluation revealed only two panels out of 209 manufactured in the given time period to exhibit this out-of-tolerance condition. Therefore, this is considered a limited incident of personnel error by Westinghouse's subvendor (Halmar) specific to these panel These errors were not identified in the Westinghouse surveillance review of the test data due to the sampling nature of this review. Since the time period of this issue, but not as a result of this apparent violation, Westinghouse subtier supplier generated documentation has been subjected to increased frequency and depth of review by Westinghouse prior to authorization for shipmen . Date When Full Comoliance Will Be Achieved CPSES is presently in complianc *
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U.S. Nuclear Regulatory Commission    ! 6 s [ MAR 3 01987
                  .-gyo . # 2 a
        .
                              ,
ATTN: Document Control Desk    i i-Washington, D.C. 20555 f, -
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          .m%.4O / . D-r, < ,.            _ . _ . . . , . . ,   %.  .      m ,,    .
SUBJECT: COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
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DOCKET NOS. 50-445 AND 50-446 INSPECTION REPORT NOS. 50-445/85-07 AND 50-446/85-05 CLARIFICATION OF SUPPLEMENTAL RESPONSE TO NOV 445/8507-01 AND 446/8505-01 REF: 1) TUGC0 letter TXX-4727 from W. G. Counsil to E. H. Johnson dated April 2, 1986 2) TU Electric letter TXX-6211 from W. G. Counsil to the NRC dated January 28, 1987 Gentlemen:
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In further review of our supplemental letter identified in Reference 2, we have determined that a clarification to the subject response is required. The supplemental request for information for Notice of Violation 445/8507-01 and 446/8505-01 (Reference 2) has been restated to aid in understanding our revised respons .
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NOTICE OF VIOLATION ITEM 1 (445/8507-01 AND 446/8505-01)
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SUPPLEMENTAL REQUEST FOR INFORMATION We find that the corrective steps taken and results achieved do not indicate that the installation of the replacement RTE-Delta Potential Transformer tiltout assembly was reinspected and accepted. Also, there is no indication that these deficiencies were documented on a nonconformance report for Units 1 and REVISED SUPPLEMENTAL RESPONSE TO ITEM 1 In our response, we noted the deficiencies involving the Unit 1 RTE-Delta Potential Transformer Tiltout assemblies were corrected using startup work permits. We fail rd to note that inspections and the acceptance of completed  l work are inherent in the startup work permit program. We ha/e confirmed these   '
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activities were accomplished as documented on the completed work permits (Z-2912 & Z-2914).
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  '*#. TXX-6349 MARCH 27, 1987
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REVISED SUPPLEMENTAL RESPONSE TO ITEM 1 - CONT'D As noted in our review of a previous NRC finding (445/8407-01), the program for managing 10CFR21 defects was not positively controlled. In some cases, and the case of the RTE-Delta 10CFR21 notice involved in this finding,-the deficiencies were corrected without nonconformance reports. Our current programs for reportability evaluation and nonconformance control are specified in procedures NE0 CS-1, " Evaluation of and Reporting of Items / Events under 10CFR21 and 10CFR50.55(e)," and NEO 3.05, " Reporting and Control of Nonconformances." Procedure NEO CS-1 requires personnel to notify the Site Coordinator upon receipt of 10CFR21 information. The Site Coordinator is required to transmit the documented information concerning a potentially significant item to the appropriate manager in the Nuclear Engineering and Operations group. Procedure NEO 3.05 states that Engineering and Operations are responsible for initiating or requesting Nonconformance Reports (NCRs),
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where required, for identified nonconformances. These programs should ensure that 10CFR21 notifications are evaluated and any required NCRs are initiate Our previous response incorrectly indicated that the Unit 2 10CFR21 deficiencies were corrected in August 1985. The work was actually completed in September 1985 and QA/QC review of the work packages was completed in April 198 Very truly yours, 1 f&lka/
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W. G. Counsil RSB/ef c- Mr. E. H. Johnson,: Region IV Mr. D. L. Kelley, RI - Region IV     t Mr. H. S. Phillips, RI - Region IV    '
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Revision as of 06:26, 24 January 2021

Partially Withheld Memo Discussing Concerns Re Insp Repts 50-445/85-07 & 50-446/85-05.Author Directed & Concurred W/ Insp Plan & Approved Insp Findings.Draft Memos,Ltrs to Util, Draft Insp Repts & Notice of Violation Encl
ML20238A048
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 08/12/1986
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To: Johnson E
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20237K807 List: ... further results
References
NUDOCS 8708200352
Download: ML20238A048 (48)
v • d • e


Text

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 \    UNf7ED STATES NUCLEAR REGULATORY COf.1 MISSION 5 / ' '

REGloN IV 

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811 RYAN PLAZA DRIVE. SUITE 1000 ARLINGTON, TEXAS M011 AUG 12 886 MENORAHDUM FOR: Eric H. Johnson, Director Division of Reactor Safety and Projects FROM: 

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Chief, Reactor Project Section 8. RP8 SULJECT: _ FOLLOWUP DN INTERVIEW WITH NRC IN'SPECTION PERSONNE - ASSIGNED STATION (CPSES) TO THE COMANCHE PEAK STEAM ELECTRIC GENER 

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REFERENCE: itRC IR N /85-07; 50-446/85-05, . dated February 3,1985 Meeting on February 25, 1986, same subject Memorandum, March 25,1986 E. H. Johnson tog dated Me ran um H. Johnson to dated As requested in your March 25, 1986, memorancum, and the later request in you June 5,1986, memorandum this documents the conce'rns related specifically to the F CPSES bruary tlRC Inspection Report No. 50-445/85-07; 50-446/85-05, dated 3, .198 inspection plan and approved the inspection findings.As the Chief CP relayed to the licensee at the conclusion of the inspection.These findirigs were 

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The items of concern regarding the subsequent handling of inspection findings were relayed to you for your consideration during discussions prior to the February 25, 1986, meeting conducted in your office. The February 25, 1986, included certain other Region IV DRSP personnel , including myself and ndin The items of concern regarding the handling of inspection again discussed in some detai Enclosures findings, my 1 through 7 to this memorandum address, as I understand the concern Enclosures 8 and 9 are draft memoranda requesting consideration comitments to these be provided requirement regarding the ASME Code requirements and applicant I appreciate your consideration regarding the items of concern and your ASME Code questions. support in obtaining the needed consideration and assistance r 

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eactor Pro ect ,

Enclosures:

Asstated(9) , Attachment: ) Draft NRC IR No. 50-445/85-07; 50-446/85-05

REGION IV 

  , A e      PARKWAY CENTRAL PLAZA BulLDING 811 RYAN PLAZA DRIVE. SulTE 1000
  "'s' e ,' ,, ' . */       ARLINGTON. TEXAS 79011
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In Reply Refer To: . Dockets: 50-445/85-07 50-446/85-05 Texas Utilities Electric Company , ATTN: M. D. Spence, President, TUGC0 Skyway Tower l 400 North Olive Street ' Lock Box 81 Dallas, Texas 75201 Gentlemen: l This refers to the inspection conducted under the Resident Inspection Program by Mr. H. S. Phillips and others during the period April 1, 1985, through June 21, 1985, of activities authorized by HRC Construction Permits CPPR-125 and CPPR-126 of the Comanche Peak facility, Units 1 and 2, and to the discussion of our findings with Mr. J. T. Merritt, and other members of your staff at the conclusion of the inspectio Areas examined during the inspection included plant status, action on previous NRC inspection findings, action on applicant identified designWithin construction these deficiencies (10 CFR Part 50.55(e) reports) and plant tour areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors. These findings are documented in the enclosed inspection repor During this inspection, it was found that certain of your activities were in violation of NRC requirement Consequently, you are required to respond to this violation, in writing, in accordance with the provision of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this lette *~' 

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l Texas Utilities Electric Company 2 Should you have any questions concerning this inspection, we will be pleased to discuss them with yo

Sincerely, . 

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Enclosures:

Appendix A - Notice of Violation Appendix 8 - NRC Inspection Report 50-445/85-07 50-446/85-05

REGION IV== NRC Inspection Report: 50-445/85-07 Permit: CPPR-126 50-446/85-05 CPPR-127 Docket: 50-445; 50-446 Applicant: Texas Utilities Electric Company (TUEC) ** l Skyway Tower 400 North 011"e Street Lock Box 81 Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 

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Inspection At: Glen Rose, Texas 

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Inspection Conducted: April 1,1985, through June 21, 1985 

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Inspectors: H. S. Phillips, Senior Resident Date Reactor Inspector Construction (pars. 1, 2, 3, 8, 9, 10, 11, 15, 16, 17, 18, and 19) J. E. Cummins, Senior Resident Reactor Date Inspector Construction (April 1 - May 10, 1985) { 

 (pars. 1, 3, and 19)    )

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Date D. M. Hunnicutt, Section Chief 

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Date Reactor Projects Branch 2 Approved: $ (pars. 1, 4, 5, 6, 7, and 19) 1 D. M. Hunnicutt, Section Chief, Date f,, Reactor Project Section B ' _ _ _ - l

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Inspection Summary Inspection Conducted April 1, 1985, thecugh June 21, 1985(Report 50-445/85-07) Areas Inspected: Routine, announced and unannounced inspections of Unit I which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), review of documentation for site dams, and review of 10 CFR Part 21 and 10 CFR Part 50.55(e) construction deficiency status, The inspection involved 77 inspector-hours onsite by . four NRC inspector Results: Within the areas inspected, three violations were identified (failure to promptly correct an identified problem with RTE - Delta Potential Transformer Tiltout Subassemblies, paragraph 3.a.; and commercial non-shrink grout was used to grout the Unit 1 reactor coolant pump and steam generator supports in lieu of Class "E" concrete, paragraph 3.b.; hydrogen recombiners out-of-specification voltage recorded on quality release document, paragraph 3.c).

Inspection Summary Inspection Conducted April 1, 1985, through June 21, 1985 (Report 446/85-05) Areas Inspected: Routine, announced and unannounced inspections of Unit 2 which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), review of documentation for site dams, review of documentation for voids behind the stainless steel cavity liner of reactor building, observation of NDE on liner plates, inspection of concrete batch plant, review of calibration laboratory records for batch plant, review of concrete laboratory testing, inspection of level C and D storage, review of reactor pressure vessel (RPV) and piping records / completed work, and review of 10 CFR Part 21 and 10 CFR Part 50.55(e) construction deficiency status, and review of violation and unresolved items statu The inspection involved 355 inspector-hours onsite by four NRC inspector Results: Within the sixteen areas inspected seven violations were identified (failure to provide objective evidence to show that concrete central and truck f mixer blades were inspected, paragraph 8; failure to issue a deficiency report { on cement scales that were out-of-calibration, paragraph 9c; failure to j translate design criteria into specifications, procedures, and drawings, l l paragraph 12a.; failure to maintain RPV installation tolerances / document v nonconformance, paragraph 12b.; failure to audit RPV specifications, j procedures, installation, and as-built records, paragraph 12d.; failure to identify a spool piece, paragraph 14b.; and failure to maintain material , I records, paragraph 14 ? 

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DETAILS Persons Contacted Applicant Personnel M. McBay, Unit 2 Reactor Building Manager . B. Ward, Gen. Supt., Civil D. Chandler, QA/QC Civil Inspector W. Cromeans, QA/QC, TUGC0 Laboratory / Civil Supervisor 

  *#J. Merritt, Assistant Project General Manager
  *#P. Halstead, Construction Site QA Manager
  #C. Welch, QA Supervisor TUGC0 (Construction)

J. Walters, TUGC0 Mechanical Engineer K. Norman, TUGC0 Mechanical Engineer J. Hite, B&R Materials Engineer G. Purdy, B&R CPSES QA Manager 

  * Denotes those present at May 10, 1985 exit intervie # Denotes those present at June 10, 1985 exit intervie The NRC inspectors also interviewed other applicant employees during this inspection perio . Plant Status Unit 1 At the time of this inspection, construction of Unit I was 99 percent complete. The fuel loading date for Unit 1 is pending the results of f ongoing NRC review l
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Unit 2 At the time of this inspection, construction of Unit 2 was approximately 74 percent complete. Fuel loading is scheduled for approximately 18 months after Unit 1 fuel loadin t I Applicant Action on Previous NRC Inspection Findings 

        ! (Closed) Unresolved Item 445/8440-02: Potential Problem with Potential Transformer Tiltout Subassemblie By letter dated June 15, 1983, Transamerica Delaval notified the l applicant of an RTE - Delta 10 CFR Part 21 report to the NRC reporting a potential problem with the primary disconnect clips of the potential transformer tiltout assembly used in the emergency diesel generator control panels at CPSES. The Transamerica Delaval l

letter also provided instructions for correcting the proble i However, the NRC inspector could not determine if the problem had l ___

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been corrected at CPSES and made this an unresolved item. The applicant determined that the problem had not been corrected and subsequently performed the recommended corrective action. The Unit I corrective action work activities were documented on startup work permits Z-2912 (train A) and 2-2914 (train B). The Unit 2 work activities are being tracked as master data base (MDB) item 3003-3 The failure to promptly correct this identified problem is an apparent violation (445/8507-01; 446/8505-01).

. (Closed) Unresolved Item 445/8416-03: Commercial Grout Used in Lieu of Class "E" Concrete The applicant determined that the use of non shrink commercial grout in lieu of the Class "E" concrete specified on drawing 2323-S1-0550 was acceptabl Design Change Authorization 21179 was issued to drcwing 2323-51-0550 accepting the use of the commercial non-shrink grou However, the failure to grout with Class "E" concrete es specified on the drawing at the time the work was accomplished is an apparent violation (445/8507-02). (Closed) Unresolved Item 445/8416-04: Hydrogen Recombiners - Out-of-Specification Voltage Recorded on Westinghouse Quality Release Document Quality Release N-41424 was revised changing the specified voltage from 10+-2V to 12+-2V which put the questionable voltage within specification limit However, the failure of receipt inspection to verify that the QRN-41424 was filled out accurately as required by Procedure QI-QAP7.2-8 is an apparent violation (445/8505-03). (0 pen) Unresolved Item 445/8432-06; 446/8411-06; Lobbin Report Described Site Surveillance Program Weaknesses During this reporting period the NRC inspector reviewed the status of this open item several times and interviewed TUEC management and site surveillance personnel. The Lobbin report stated that the scope and objectives of the site surveillance program were unclear, lacking both purpose and directio There is no specific regulatory requirement to have a surveillance program; however, TUEC committed to have a surveillance program and has established procedures to implement such a program as a part of the 10 CFR Part 50, Appendix B, QA program. This extra effort is a strength; however, the NRC inspector also observed, as did the Lobbin Report, that the surveillance program lacks both purpose and direction to be effective and complimentary to the audit and inspection program Since the TUEC audit group is not located on site, the TUEC surveil-lance program on site takes on added significanc This item was discussed with the TUEC site QC manager who described 1 a reorganized site surveillance function and changes that have occurred. New procedures which describe this organization's duties and responsibilities are forthcoming.

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   -5-TUEC has elected to defer responding to the violations pertaining t the audit function in NRC Inspection Report 445/84-32; 446/84-11, but rather to have the Comanche. Peak Response Team'(CPRT) respond.to this repor^t and other QA matter The surveillance issue is closely tied to the audit deficiencies in NRC Inspection Report No. 445/84-32; 50.446 84-11. This. item will remain open pending the review and implementation of the CPRT action pla A special point of interest , ,

will be how audits and surveillance work together to evaluate the control of all safety related activities on site to essure quality especially the overview of quality control effectivenes . Document Inspection of Site Dams The NRC inspector reviewed documents describing the inspection activities performed on the Squaw Creek Dam (SCO) and the safe shutdown impoundment (SSI) for impounding cooling water for the two units at CPSES. The  ! 

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purpose of the SCO is to impound a cooling lake for CPSES. A secondary reservoir-(SSI) is formed by a channel connecting the SCO impoundment to the SS Three documented inspections have been performed since 198 The inspections were: Relevant data for'SCD is contained in Phase I Inspection, National Dam Safety Program, Squaw Creek Dam, Somervell County, Texas, Brazos River Basin, inspection by Texas Department of Water Resource Date of Inspection: June 10, 198 Inspection on August 25, 1982, by registered professional engineers from Mason-Johnston & Associates, Inc., and Freese & Nichols, In I Inspection on September 19, 1984, by a registered professional engineer from Mason-Johnston & Associates, In The inspection activities consisted of visual inspections by inspection teams that included accompanying Texas Utilities Service, Inc. (TUSI), and Texas Utilities Generating Company (TUGCO) representative Photographs were taken as a part of the documentation. The data for the 

. piezometer observations and the data for the surface reference monuments were reviewed by applicant personnel and Mason-Johnston engineer No items of significance were observed or reported by these inspection teams. Slight erosion areas were observed and reporte A cracked area on the service spillway upstream right bridge seat was observed by the inspection teams and continued monitoring of this area was recommended by Mason-Johnston and Associate No signs of cracks, settlements, or horizontal movement at any location within the SCO or the SSI were l reported.

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The NRC inspector reviewed the applicant's records and the Mason-Johnston inspection reports. These documents indicated that the SCD and SSI were structurally stable and that the applicant was performing inspection activities to maintain the structural integrity of these dam The state of Texas requires periodic inspections of these dams (principally the SCO) due to inhabited dwellings downstream. The applicant has met these inspection requirement . . No violations or deviations were identifie . Voids Behind the Stainless Steel Cavity Liner in Unit 2 Reactor Building The NRC inspector reviewed applicant records, including NCR C-82-01202; NCR C-1784, Rev. 1; NCR C-1784, Rev. 2; NCR C-1766, Rev. 1; NCR C 1791, Rev. 1; NCR C-1824, Rev. 1; NCR C-1824, Rev. 2; Significant Deficiency Analysis Report (SDAR) - 26, dated December 12, 1979; DCA-20856; and Gibbs and Hill Specification 2323-55-18. The review of these records and documentation and discussions with various applicant personnel indicated the following: Structural concrete was placed in Unit 2 reactor building at elevation 819 feet 6-3/4 inches to 846 feet 6 inches on June 21, 1979. This concrete was placed adjacent to the stainless steel liner walls. The concrete forms for this pour were not removed until October 1979 due to subsequent concrete placements for the walls to elevation 860 feet 0 inches. When the forms were removed, honeycombs and voids were observed by applicant personne The applicant's review of the extent of unconsolidated concrete resulted in the issuance of SDAR-26 on December 12, 1979: Investigations were begun and Meunow and Associates (M&A) of Charlotte, North Carolina, were contracted to perform nondestructive testing on in place concrete. M&A performed these tests on a two foot grid pattern on the compartment and liner sides of all four steam generator (SG) compartment walls. The selected test locations did not include the locations where the voids were later found to be located; therefore, the voids were not detected during the M&A testin In August 1982, preparations were made to pour the concrete annulus around the reactor vessel. When the expanded metal formwork was removed from the reactor side of the compartment walls, voids were observed and NCR C-82-01202 was prepare DCA 20856 was prepared as a procedure to repair the void area. DCA 20056 indicated that the voids were not extensive (a surface area of about 28 square feet by i 8 inches maximum depth) and that the repair procedure assured that the total extent of voids had been identified. One half (0.5) of a { I cubic yard of concrete was used to complete the repairs as indicated on grout pour card 26 The applicant's review and evaluation of the gird pattern and a comparison of SG compartments 2 and 3 to 1 and 4 indicated that voids did not exist in SG compartments 2 and 3. The review of test girds extended down to elevation 834 feet, which is the floor elevation of the line The liner walls of SG compartments 1 and 4 _________-___________m _ 

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  -7-were not tested at elevation 834 feet, but at elevation 836 feet  !

which is above the area of the identified voids. No testing was j done on the liner side of the area of the ;ide below elevation 836 

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feet. The program also included removal of c 7ch x 2 inch plugs f I from the stainless steel liner at locations where test indications raised questions concerning the concrete. The inspections of the concrete by applicant personnel after the plugs were removed confirmed that there were no additional unconsolidated concrete - areas (voids).

. The applicant removed stainless steel liner plates from three areas (one arec about 1 foot by 1 1/2 feet and two areas about 3 feet by 1 foot, excavated or chipped to sound concrete, and cleaned the concrete surface area. One and one quarter inch (11/4) diameter probe holes and grout access holes were drilled in the liner plates to determine the extent of and to assure full definition of the void are Air access holes were drilled in the stainless steel liner plates to assure that grouting would be accomplished in accordance with the procedur The procedure (DCA-20856) specifed that the grout was to be cured for 28 days or until the grout reached a compressive strength of 4000 psi. Repairs to the liner plates were specified in DCA-20856 and G&H Procedure 2323-55-1 DCA-20856 required that under no circumstances was cutting of the liner across weld seams, across embedded weld plates, or into leak chase seal welds or drilling through the liner at leak chase channels, embeds, or weld seams permitte Documentation review indicated that DCA-20856 was adhered to and that no cutting or drilling occurred in prohibited location No violations or deviations were identifie . 6. Nondestructive Testing Observations of Liner Plates in Fuel Transfer Canal The NRC ir;pector observed portions of non-Q liquid penetrant examinations (PT) being performed on liner plate welds following re-installation of the liner plates in the areas of the fuel transfer canal removed for inspection and repair of the concrete. The inspector performed the PT on the welds as required by the repair package and the procedure (QI-QP-11.18-1, 

" Liquid Penetrant Examination"),. Scattered weld porosity was identified by the inspection. The porosity was ground out and a repeat PT was performed. The final inspection is scheduled to be performed by QC inspection personnel. The liner plate areas to be inspected by PT were identified in DCA 2085 No violations or deviations were identifie I

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     -8-7. Rebar Placement and Cadweld Splice Observations and Records The NRC inpector reviewed the rebar placement and Cadwell Splice activities associated with the Unit 2 containment (reactor butiding)

closur Calibration of Tensile Tester The NRC inspector observed the' calibration of the Tinus-Olson , Universal Testing Machine (Model Number 600-12 Identification Number M&TE-784)'on April 2 and May 7, 1985. The machine was calibrated just prior to performing tensile testing of cadweld splices and ' subsequent to completion of' tensile testing each day that tensile testing was performed. The machine calibration date for April 2, 1985, prior to start of tensile testing was observed by the NRC inspector and recorded as follows: Nominal load Calibration Reading Error Error Remarks (lbs) (1bs) (lbs) % 0 0 0 0 0 machine on 4/2/85 100,000 99,750 +250 +0.25 200,000 199,600 +400 + ,00 299,450 +550 +0.18 350,000 350,300 -300 -0.08 400,000 401,200 -1200 -0.03 500,000 501,350 -1350 -0.27 600,000 602,450 -2450 -0.40 The NRC inspector reviewed calibration data for March 4, March 8, April 2, April 3, April 30, and May 7, 1985. All calibration data met within the +/- 1% accuracy requirement specified by Calibration Procedura 35-1195-IEI-37, Revision 3, dated March 11, 1982. The reference standards were identified as follows: ID N Manufacturer Calibration Due Date RS-75 BLH Electronics January 27, 1987 RS-7 BLH Electronics January 27, 1987 v Observation of Cadweld Splice Tensile Testing (1) Qualification Tensile Testing On April 2, 1985, the NRC inspector observed the following tensile testing of cadweld splices for cadwelder qualification: EBD Q8, GBH Q1, GBH Q2, GBV Q1, BFD Q4, BFD Q3, BFH Q4, GAH Q1, GAV Q1, and GBV Q Each of the above qualification cadweld splices was tensile tested to 400,000 pounds (100,000 psi) and met the requirements stated in the procedur _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _

_ -- _ _ _ _ _ _ _ _ _ _ _ _ 

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 (2) Production Tensile Testing        ,

r ? The NRC inspector observed the tensile tester calibrations and /,-i' the following production cadweld splices tensile. testing on i May 7,1985: FXD 3P, FYD 4P, FYD 8P, FRD 87P,.and FUD 6 l 

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           , .e Each of the above production cadweld splices was tested to 400,000 pounds (100,000 psi)and met the requirements stated in
            'h
            '
            '

the procedur * ' 

 (3) Installation of Production Cadweld Splices        I The NRC inspector observed installation of rebar and cadweld       ,

splices at frequent ir,tervals (five or more observations per ,/ week during the weeks of April 8 and 15; May 6, 13, 20, and 27; and June 3, 1985). The rebar installation for the Unit s closure was performed in the area identified as elevation' 80 , feet to elevation 875 feet and azimuth 300 degrees to 335 : . 

            '.,

degrees. The installation activities observed included'rebar spacing, location of cadwelds, observation of selection and \ removal for testing of cadweld splices for testing, and determination of location of rebars and cadwelds for the as-built drawing (4) Documentation Reviewed e The NRC inspector reviewed the following documentation for the rebar placement and cadwelding for the Unit 2 containment (reactor building) closure area: Drawings DCAs NCRS 2323-5-0785, Re , Rev. 1 C85-200294 2323-5-0786, Re C85-200339 Re , 2323-S1-500, Re C85-200355, Re , Re , Re (Sheets 1-7) i 2323-52-508, Re ?2 2323-52-506, Re l

l No violations or deviations were identifie . Concrete 8atch Plant Inspection l The NRC inspecor used a nationally recognized checklist to inspect the  ! concrete production facilities This list included the specific ' characteristics for the foll ning areas: (1) material storage and handling of cement, aggregate, water and admixture, (2) batching equipment scales, weighing systems, admixture dispenser, and recorders, (3) central mixer (not applicable because it had been dismanteled), (4) ticketing system, and (5) delivery syste l 

      . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - -   ~

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y ?The current batching is a manual operation since almost all concrete has 

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  -been placed. The central mixer was dismanteled and removed from site two
,
 , or three years ago when concrete placement was virtually complete ~'

Presently, the backup batch plant (which was a backup system for the 

,

central mixer) is in operation to complete the remaining concrete 

/~   plac qents. This batch plant is in good condition and complied with the f '
 ;

subject dmcklist except for one area The NRC inspector inspected the inside of one of three trucks used for . . mixing; concrete (that is, the batch plant dispenses the correct weight of materials as required by the specific design mix numbers and the truck then mixes the batch to be placed.) The blades inside the truck are 

,

subject to wear and should be checked at a reasonable frequency. The Brown & Rost (B&R) representative responsible for checking the blades in accordance witn B&R Procedure 35-1195-CCP-10, Revision 5, dated December 4, 1978, was asked for evidence that the blades had been checked for wear on a quarterly basis and it was found that there was no record of i such checks dating back to 1977 when they were initially checke Procedure CCP-10, paragraph 3.10 " Truck Mixing", is silent on blade wear but Section 3.11 infers that the blades should be checked for both central and truck mixing. The inspection of both central and truck mixing blades was not documented, although the B&R representative stated that the mixing blades were periodically inspected and laboratory testing would have probably indicated if there was a problem with the mixing blade Strength and uniformity tests have consistently been within the acceptable range indicating that concrete production was acceptable even though mixing blade inspection was not documente Otherwise, the condition of the inside of the truck was satisfactory as the drum and charging / discharging were clea The water gage and drum counter were in good conditio This failure to follow procedures is a violation of 10 CFR 50, Appendix B, Criterion Subsequent to the identification of this violation, the blades were checked for wear and blade wear was presently within allowable limits (445/8507-04; 446/8505-02).

No other violations or deviations were identifie Calibration Laboratory for Batch Plant l The NRC inspector obtained batch plant scale numbers from tags which l indicated that the scales had been calibrated and were within the calibration frequenc Cement (MTE 779), Water (MTE 766), admixture scale (MTE 764), and aggregate (MTE 780) were reviewed. The scales had been periodically calibrated since the batch plant was activate The records were adequate except as follows:  ! Scales MTE 766 records do not clearly differentiate between the required accuracy of the scale and the digital readou * ., i 

.
      -. -
   -11- Scales same MTE 779 and 780 records show various accuracy ranges for the scale; i.e., MTE 779 (SN749687) records the following: report dated January 1976 gives EE; report dated July 1976 gives 1% while report October 1976 gives +/- 0.2%.
      .

The above items are unresolved pending further review of the licensee actions regarding these records during a subsequent inspection (445/8507-05; 446/8505-03). The calibration appeared to be prope * 

      * * Records July for scales 16, 1975, MTE 779 records contained B&R memo IM-1108 d which described a nonconforming conditio This condition affected the water and cement scales causing a 24-48 pound deviation during the calibration tes The memo stated that the condition was corrected and the scales were then calibrated; however, no deficiency report was written as required by B&R Procedure CP-QCP-1.3, " Tool and Equipment Calibration and Tool Control" dated July 14, 1975, and CP-QAP-15.1, " Field Control of Nonconforming Items," dated July 14, 1975. As a result there is no evidence if concrete that corrective production action included was adversely affected.an evaluation to determine This failure to assure that a nonconforming condition was evaluated is a violation of Criterion XV of 10 CFR Part 50, Appendix (445/8507-06; 446/8505-04).

1 Concrete Laboratory Testina TUGC0 procedure QI-QP-11.1-1, Revision 6, was compared with ASME Section III, Division 2, Subsections 5222, 5223 and 5224 to assure that each ASTM testing requirement was incorporated into the procedur The NRC inspector inspected the testing laboratory equipment and found the test area was in good condition and each piece of equipment was tagged with a calibration sticker which showed it to be within the required calibration frequenc test requirements and equipment. Test personnel were knowledgeable of Thefollows: as NRC inspector witnessed field tests performed by laboratory personnel Date Truck No. Mix No. Ticket N Air Content (%) Slump Temperature (in.) _F 6/3/85 RT-41 925 64013 Req 8.2-1 NA 70 max Mea 8.7- NA 57 6/3/85 RT-35 128 64014 Req 5.0- max 70 max < Mea .25* 57 

* Truck was rejected by quality control but was later accepted when second slump reading came into required rang E
*
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i-12- -- The following laboratory equipment was checked and found to be within calibration: Forney Compression Tester, MTE 3031; Temperature Recorder ; MTE 3013 and 3014; Unit Volume Scale, MTE 1053; Pressure Meters MTE i 3000B, 3002 and 3004; Sieves MTE 1286, 1239, 1272, 1274, 1136A, 1156, 1094, 1093, 1095, 1178, 1179, 1300 and 1180; Aggregate scales, MET 1058 and 1067; and 2" grout mold MTE 111 The following test records for placement number 201-5805-034 were reviewed: (1) concrete placement inspection, (2) concrete placement - 

        .

summary and, (3) unit weight of fresh concret l No violations or deviations were identifie . Inspection of Level C and D Storage I The NRC inspector inspected all laydown areas where ciping, electrical conduit, cable, and structural reinforcing steel were store These materials were neatly stored outside on cribbing in well drained areas which allowed air circulation and avoided trapping wate This met the Level "D" storage requirements of ANSI N45. The electrical warehouse contained miscellaneous electrical hardwar This building was required to be fire and tear resistant, weathertight, and well ventilated in order to meet Level "C" storage requirements. This warehouse was well kept and met all requirements except for a lock storage area located upstairs at the rear of this building (electrical termination tool room). Two minor problems were identified and the warehouse personnel initiated action to correct the The first problem noted was that a box of nuclear grade cement was marked 

  " shelf life out of date" but it had no hold ta The box was subsequently tagged with Nonconformance Report (NCR) E85-200453 af ter being identified by the NRC. During discussions with the warehouseman, the NRC determined that engineering told the warehouseman to mark the material and lock it up, but did not tell him to apply an NCR or hold ta TUEC should determine if engineering is aware of nonconforming material controls and provide training if this is other than an isolated instance. Also, the NRC inspector noted a very small leak in the roof above the electrical termination tool room. This lesk was in an area that did not expose hardware to moisture. The roof is currently being repaire The millwright warehouse storage area was inspected; however, only a small number of items or materials were stored in this area. The overall storage conditions in this area met or exceeded Level "C" storage requirement No violations or deviations were identifie . Reactor Pressure Vessel and Internals Installation - Unit 2 This inspection was performed by an NRC inspector to verify final placement of the reactor pressure vessel (RPV) and usernals by examining the completed installation and inspection record I
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a. Requirements for Placement of RPV Requirements for placement of the RPV to ensure proper fit-up of all other major NSSS equipment are in Westinghouse Nuclear Services , Division (WNSD) " Procedure for Setting of Major NSSS Components", ) Revision 2, dated February 13, 1979, and " General Reactor Vessel Setting Procedure" Revision 2, dated August 30, 1974. The NRC ' inspector reviewed the following drawings, which were referenced in the RPV operation traveler, to verify implementation of WNSD recommendation . j o WNSD drawing 1210E59 " Standard - Loop Plant RV Support Hardware Details and A,sembly" o WNSD drawing 1457F27 " Comanche Peak SES RCS Equipment Supports 

  - Reactor Vessel Supports"    .

l o CE drawing 10773-171-004 " General Arrangement Elevation" l o CE drawing 10773-171-005 " General Arrangement Plan" l Neither site prepared installation drawings nor specifications (which implemented the WNSD recommended procedures) were available and the drawings examined did not show certain specific installation criterion such as centering tolerances, levelness tolerances and clearance between support brackets and support shoe The lack of engineering documentation did not provide full control of the action and would allow changes to installation criteria important to safety to be made without complying with established change procedure This is considered a violation of 10 CFR 50, Appendix B, Criterion III (446/8505-06).

b. Document Review The NRC inspector reviewed B&R Construction and Operation Traveler  ; No. ME79-248-5500 which described the field instructions for ] installation of the Unit 2 RP Requirements recommended by WNSD 1 procedures were implemented in the traveler. Worksheets attached to the traveler showed the RPV to be centered and leveled within the established tolerances, Traveler operation 19 required verification of a 0.020 to 0.005 inch clearance between the support bracket and ) support shoe, after applying the shim plates. Change 5 subsequently ] changed the clearance to a 0.015 to .025 inch clearance. The installation data reflected in attachment 3B of the traveler indicated an as-built clearance of 0.012 to 0.026 inch which exceeds both the original and revised tolerances. This condition was accepted on the traveler based on Westinghouse concurrence, and there were neither nonconformance reports nor documented engineering evaluations to determine if the condition was acceptable. This failure to document nonconforming conditions and engineering deviations is a violation of 10 CFR 50, Appendix B, Criterion XV (446/8505-07) The NRC inspector reviewed the following receiving records for RPV hardware and found them to be in order:

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  .'o Report No.14322 for 54 each closure studs, closure nuts, and closure washers o Report No. 09507 for vessel S/N 11713, Closure Head 11713 an Rings o Deviation notices and corrective action statements The NRC inspector reviewed the following completed travelers'for   -

internals installation and found them to be satisfactory; o ME-84-4641-5500, " Assemble Upper Internals" o' ME-84-4503-4000, " Install and Adjust Roto Locks o ME-81-2145-5500, "Retorque UI Column Extension" o RI-80-385-5500, " Transport and Install Lower Internals" o ME-84-4617-5500, " Repair Lower Internals" o ME-84-4640-5500, " Assemble Lower Internals" Visual Inspection At this time, visual inspection of the internals by the NRC inspector was not possible, and inspection was limited on the vessel placement to a walk-around beneath the vessel to inspect the azimuth markings and for construction debris between the vessel and cavity. No problems were identified in this are Records of QA Audits or Surveillance

The NRC inspector requested TUGC0 QA audits or surveillance performed by TUGC0 of the Unit 2 RPV installation. TUGC0 did not make available any audit or surveillance reports of specifications for placement criteria, placement procedures, hardware placement, or i as-built records. Failure to perform audits or surveillance of RPV specifications, procedures and installation is a violation of i 10 CFR 50, Appendix B, Criterion XVIII (446/8505-08).

No deviations were identified; however, three violations were identified and are described in the above paragraph . Reactor Vessel Disorientation On February 20, 1979, the applicant reported to the NRC Resident Inspector that a design error had resulted in the reactor support structures being placed in the wrong position on the reactor support pedestal such that the reactor would be out of position by 45 degree Initially, Unit 2 was to be a mirror image of Unit 1, however, a design change was initiated to permit identical components for both units. The I - - _ - _ _ _ _ _ _ .__ _ _ _ _ _ 

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design change was implemented for the reactor vessel, but not for the pedestal support location The problem was not considered by the applicant to be reportable under provisions of 10 CFR Part 50.55(e) since the error could not have gone undetecte The problem was reported to the NRC Office of Inspection and Enforcement on February 22, 1979, and during a March 27, 1979, meeting in Bethesda, Maryland, the applicant presented the proposed redesign and rework , procedures for relocating the pedestal support No unresolved safety concerns with the repair were identified at the meetin During this inspection the NRC inspector reviewed various documentation relative to the disorientation problem, including design changes and the } construction traveler which implemented the repai The following documents were reviewed: o NRC Inspection Report 50-446/79-03 o NRC Inspection Report 50-446/79-07 o NRC Inspection Report 50-446/79-13 - o TUSI Conference Memo, dated March 1, 1979, H. C. Schmidt to S. Burwell (NRC Licensing PM) o TUGC0 letter TXX-2980, dated April 30, 1979, to W. C. Seidle o NRC letter to TUGC0 dated May 29, 1979 o DCA 3872, Revision 1, dated February 28, 1979, Subject: Rework of Structure for Placement of the RPV Support Shoes I o DCA 4122, dated March 22, 1979, Subject: Replacement of Rebar for RPV Supports o Construction Traveler CE79-018-5505, dated March 14, 1979, Subject: Rework of Reactor No. 2 Cavity - New RPV Support Locations o Grout Replacement Cards No. 007,008, 009, 010, 014 and 015, various dates, Subject: Replacement of Grout around Rebar for Repair of RPV 

 ,, Support Shoes o Various Inspection Reports for Grout Properties and Application for RPV Support Shoes No violations or deviations were identifie . Reactor Coolant Pressure Boundary (RCPB) Systems The inspection was performed to verify: the applicants system for preparing, reviewing, and maintaining records for the RCP8 piping and components; that selected records reflected compliance with NRC requirements and SAR commitments for manufacture, test and installation

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   -16-of items; and that as-built hardware was adequately marked and traceable to records. The following items were randomly selected and inspected:

l Pressurizer Safety Valve - This item was inspected to the commitment-stated in FSAR, Table 5.2-1 which includes ASME Section III, 1971 I edition through winter 1972 addenda. Valve S/N N56964-00-007, which is installed in the B position was inspected. The following records j were reviewed: 

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o QA Receiving Inspection Report No. 21211 < o Code Data Report Form NV-1 o Valve Body CMTR The valve was in place, however, installation had not been completed; therefore, the hardware installation inspection consisted of verifying that the item was traceable to the record CVCS Spool Piece 3Q1 - Requirements for this item are stated in ASME, .Section III,1974 edition through summer 1974 addenda, which is the commitment from the FSAR, Table 5.2-1. The item was field fabricated from bulk material and installed in the CVCS with field welds number 1 and 3 (ref. BRP-CS-2-RB-076). The following records were reviewed: o B&R Code Data Report o Field Weld Data Card o NDE Reports o QA Receiving Reports (for bulk order) o Certified Material Test Report (CMTR) The installed spool piece was inspected for weld quality and to verify that marking and traceability requirements had been me The item had been marked with the spool piece number (3Q1) and the B&R drawing number, however, marking of the material specification number and type, heat code, or other means of traceability could not be found. In respect to material requiring a CMTR, (nominal pipe size greater than 3/4 inch) NA-3766 requires marking with the applicable specification and grade of material and heat number or heat cod When material is divided, the identification marking is required to be transferred to all pieces. This failure to identify material marking is a violation 10 CFR 50, Appendix B, Criterion VIII (446/8505-09). Loop 3 RC Cold Leg - Requirements for this item are stated in ASME, , Section III, 1974 edition through summer 1974 addenda, which is the commitment from the FSAR, Table 5.2- This piping subassembly ,

c_ , 

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     -r-17-consists of a 27.5 inch cast pipe with a 22 degree elbow on the reactor end, a 10 inch 45 degree nozzle, a 3 inch nozzle, and three 2 1/2 inch thermowell installation bosses. The following records were reviewed for the subassembly; o QA Receiving Inspection Report No. 12389 o Westinghouse Quality Release (QRN 47523)  -

d "o ' Code Data Report Form NPP-1 o 27 1/2 inch line CMTR o 3 inch nozzle CMTR o Field Weld Data Cards o NDE Reports (1) The NRC inspector determined that CMTR's were required by the code but were not available for the following items: 

 . 22 degree elbow
 .

10 inch 45 degree nozzle 

 .

2 1/2 inch thermowell bosses This failure to maintain retrievable records is a violation of 10 CFR 50, Appendix B, Criterion XVII (446/8505-10).

(2) Sandusky Foundary and Machine Company test report for the cold leg pipe certifies that material meets requirements of ASME Section II, 1974 editions through winter 1975. Southwest Fabrication and Welding Company code data report NPP-1 Form certified that the cold leg subassembly met requirements of ASME Section III, 1974 edition through winter 1975. The FSAR commitment is ASME Section III, 1974 edition through summer 1974. This discrepancy is unresolved pending the applicant's evaluation to determine if material nonconformances exist (446/8505-11).

(3) The cold leg NPP-1 Form stated that no hydrostatic (Hydro) test had been performe In discussions with Westinghouse and B&R personnel, the statement was made that it is normal practice to defer the partial hydro test until the whole system is hydro tested. B&R Procedures CP-QAP-12.1 and CP-QAP-12.2 describe requirements for the tes ASME, Section III, (NB-06114(a)) states in respect to the time of testing piping subassemblies, that the component test, when conducted in accordance with the requirements of NB-6221(a) shall be acceptable as a test for piping subassemblie NB-6221(a) om m ss __n-mo_ statesm__m that completed components shall be subjoeBod e - - - -- - -

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I-18- --- -! Paragraph (b) of NB-6221 requires pressure testing of all pressure retaining components that are within the boundary protected by overpressure protection devices and paragraph (c) permits substitution of the system test for component test. It is not evident that the system test substitution was permitted for pipe subass'embly since NA-1200 makes a distinction in the 

 . definition between components and piping subassemblies. NB-6115 states that pressure testing of components shall be performed prior to initial operation of a system and that the Data Repor .

Form shall not be completed nor signed by the code inspector and i the component shall not be stamped until the component manufac- 

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turer has conducted the hydrostatic pressure test. Additionally, NA-8231 prohibits the application of a Code stamp prior to hydrostatic test regardless of whether the test was parformed prior to or after installation of the ite The NPP-1 Form for the cold leg had been completed and signed by the manufacturer and Authorized Nuclear Inspector (ANI). An ASME Code Stamp had also been applied to the item. The NRC inspector observed that a hydrostatic test had not been performed and was noted on the NPP-1 for The above items are unresolved pending clarification of code requirements by NRC headquarters (446/8505)-12).

(4) Since the cold leg pipe subassembly had not been pressure tested prior to installation, the NRC inspector reviewed the procedures and hydro test data applicable to Unit 1, since Unit 2 hydro had not been completed. Requirements for the tests were presented in Procedures CP-QAP-12.2, " Inspection Procedure and Acceptance Criteria for ASME Pressure Testing" and CP-QAP-12.1, 

 "ASME Section III Installation, Verification, and N-5 Certifi-cation." Procedure CP-QAP-12.1 requires that a data package to be used in the test, be prepared with the test boundary and the additional following data shown:

o Base metal defects in which filler material has been added, and the depth of the base metal defect exceeds 3/8 inch or 10% of the actual thickness, whichever is less, o Untested vendor performed piping circumferential weld o Approximate location and material identification and description for permanent pressure boundary attachment with applicable support number reference o Weld history, which shall reflect weld removal and/or weld repai _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . 

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The completed hydro data package (PT-5501) for Unit 1, loop 3 . cold leg was reviewed for compliance with the above requirements. Drawing No. BRP-RC-1-520-001 had been used to annotate the test boundar A handwritten statement on the , 

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drawing indicated: "No major base metal repairs could be located" and "No hangers with weld attachments could be located." Welds performed by the pipe subassembly vendor, including the 22 degree circumferential weld and the penetration fittings had not been identified. The following . items were unresolved regarding the adequacy of the hydro test: o Was the determination of no major base metal repairs based on a visual inspection or on a review of vendor and site 

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inspection and repair records? o Was the shop circumferential weld attaching the 22 degree elbow to the pipe assembly inspected during the test? If so, where is the inspection identified? o Procedure CP-QAP-12.1 does not require identification of welds for penetrations into the pipe assembly and they were not identified on the drawing. Were those welds inspected? If so, where is the inspection documented? The above issues will remain unresolved pending further evaluation by the applicant (44S/8507-07; 446/8505-13). Personnel Qualifications - Personnel who had performed selected tasks were identified during inspection of installation records. Training and experience records for the personnel were reviewed to verify that employee qualifications and maintenance of records were current and met requirement Names or codes for five welders and two NDE examiners, who had performed tasks during installation of the items being inspected, were identified and their qualification records i reviewed. There were no questions in this area of the inspectio . Special Plant Tours (Unit 1 and Unit 2) On May 23, 1985, the NRC inspector conducted a tour of selected areas of Unit 1 and Unit 2. The group consisted of one NRC inspector, two NRC Technical Review Team (TRT) representatives, two allegers, and several TUEC representative The TUEC representatives tagged each area where a deficiency was alleged. With the alleger's consent, a tape recorder was also used to note locations and describe any alleged deficiencies. The < 

         !

allegers indicated that they had identified all deficiencies during the tour and all other deficiencies that they had knowledge The NRC TRT is l 

         >

analyzing this information and will decide whu Action, if any, should be take ( During this tour the NRC inspecter independently identified a l 

         '

questionable practice in that the top of the the pipe chase at the north end of room 88 in Unit 1, safeguards building had two large stickers which

__ P ' F l 

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i-20- stated that areas on the wall were reserved.for pipe hangers GHH-51-1-5B-038-006 and R1(?)1-087-X1 These stickers were dated 198 ]; It was ??? evident whether hangers missing or none were needed in these locations and the reserve tags were not removed. TUEC representatives were unable to. answer the question immediately. This item is unresolved 3 pending further review during a subsequent inspection. (445/8507-07; .] l 446/8505-05).

No violations or deviations were identifie . , 1 Routine Plant Tours (Units 1 and 2) At various times during the inspection period NRC inspectors conducted general tours of the reactor building, fuel building, safeguards building, electrical and control building, and the turbine buildin During the tours, the NRC inspector observed housekeeping practices, preventive maintenance on installed equipment, ongoing construction work, and discussed various subjects with personnel engaged in work activitie l I No violations or deviations were identifie . Review of Part 21 and 10 CFR 50.55(e) Construction Reports Status The NRC inspector reviewed all reports issued to date to assure that NRC and TUEC status logs were complete and up to date. A total of 183 reports have been submitted to date. This inspection period one Part 21 report on Diesel Generator Oil Plugs and two 10 CFR 50.55(e) reports on the Equipment Hatch Cover and SA106 Piping (light wall; were submitte No violations or deviations were identifie , 1 Review of Violation and Unresolved Item Status The NRC inspector reviewed all violations and unresolved items reported to date to assure that NRC and TUEC status logs were complete and up to In addition, a trend date. Two hundred nineteen items were reviewe analysis of NRC findings was performed to generally determine how many findings could be broadly classified under each criterion of 10 CFR, Part 50, Appendix The frequency of findings showed broad and general trends under the following criteria: II. QA Program; III. Design Control; V. Instructions, Procedures and Drawings; VII. Control of Purchased Material, Equipment and Services; IX. Control of Special Processes; Inspection; XI. Test control; XIII. Handling Storage and Shipping; XVI QA Records; and XVIII Audits. The most significant Also, a trends were number of violations Criterion III, V, VII, IX, X, and XVII occurred with respect to 10 CFR 50.55(e) item These findings mainly pertained to Unit 1 and related closely to trends identified by the NRC Technical Review Team TRT. These trends will be considered during followup on TRT findings. Also, Unit 2 inspection emphasis will consider these trends during future inspection No violations or deviations were identifie _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ _ 

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19. Exit Interviews The NRC inspectors met with members of the TUEC staff (denoted in paragraph 1) on May 10 and June 10, 1985. The scope and findings of the inspection were discussed. The applicant acknowledged the findings.

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     -

) In Reply Refer To: Dockets: 50-445/85-07 50-446/85-05 Texas Utilities Electric Company ) ATTN: M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street ' Lock Box 81 Dallas, TL-as 75201 Gentlemen: This refers to the inspection conducted under the Resident Inspection Program by Mr. H. S. Phillips and others during the period April 1, 1985, through June 21, 1985, of activities authorized by NRC Construction Permits CPPR-125 and CPPR-126 of the Comanche Peak facility, Units 1 and 2, and to the discussion of our findings with Mr. J. T. Merritt, and other members of your staff at the conclusion of the inspectio Areas examined during the inspection included plant status, action on previous NRC inspection findings, action on applicant identified design construction deficiencies (10 CFR Part 50.55(e) reports) and plant tours. Within these . k areas, the inspection censisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors. These findings are documented in the enclosed inspection repor During this inspection, it was found that certain of your activities were in violation of NRC requirement Consequently, you are required to respond to this violation, in writing, in accordance with the provision of Section 2.201 l of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal  ! Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this lette I l RPB2/8 C/RPB2B C/RPB2 NRR HSPhillips/dc DHunnicutt VNoonan 

/ /85 / /85  8 / /85 l

_ _ _ _ - _ e ** e 

     .:- . )

Texas Utilities Electric Company 2 Should you have any questions concerning this inspection, we will be pleased to discuss them with yo

Sincerely, _ 

    .  :
      ..

eacto Project Branc 

   \

Enclosures: Appendix A - Notice of Violation Appendix B - NRC Inspection Report 50-445/85-07 50-446/85-05 cc w/ enclosure: 

    *

Tsxas Utilities Electric Company Skyway Tower 400 North Olive Street L ck Box 81 Dallas, Texas 75201 - 

      {

Texas Utilities Electric Company ATTN: J. W. Beck, Vice President Skyway Tower 400 North Olive street Ltck Box 81 Dallas, Texas 75201 bec distrib. by RIV: TEXAS STATE DEPARTMENT OF HEALTH bec: to DMB(IE01) bec distrib by RIV: SRPB1 * Resident Inspector OPS 

"RPB2  * Resident Inspector CONS  .

R. Martin, RA * D. Hunnicutt, Chief, RPB2/B 6C. Wisner, D/DRSS PA0 V. Noonan, NRR S. Treby, ELD MIS SYSTEM RIV File Ju:nita Ellis Renea Hicks CD. Weiss, LFMB (AR2015) 

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