ML20238A616
| ML20238A616 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/23/1986 |
| From: | Renee Taylor NRC |
| To: | Westerman T NRC |
| Shared Package | |
| ML20237F760 | List:
|
| References | |
| NUDOCS 8708210026 | |
| Download: ML20238A616 (39) | |
Text
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MEMORANDUM FOR:
T. F. Westerman, Chief Reactor Safety Branch FROM:
R. G. Taylor, Reactor Inspector P
-R SU8 JECT:
TO JOHNSON MEMORANDUM OF JANUARY 13, 1986 g
The following addresses the enclosure to subject memorandue: :
StructuralSteelSupports(480638,550638,55065). These procedures were intended to address building and component steel supports. With the exception of the main reactor components CPSES design uses almost no steel as either building floor supports or for component supports. These elements are all reinforced concrete. Little inspection was warranted in 78/84 and I doubt if its changed.
Records on components (50055-51055). Statement that additional inspection needed is perhaps correct, however, the September 15,_1981, version of MC2512 with its prioritization scheme was implemented shortly after issuance. This scheme sharply. curtailed all records review aspects of our construction l
inspections if only priorities 1 and 2 were implemented. I believe that Mr. Phillips is seeing the results of that decision.
With the exception of mid-tem QA, virtually all of the GA reviews discussed in this enclosure did not exist until 1980 and then were stronply deemphasized
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in the prioritization scheme issued in 1981. Rightly or wrong y, until about 1982, the emphasis in MC2512 was the end result of the QA program; i.e., was the
' finished product acceptable rather than at extensive reviews of program and resulting paperwork.
There has never existed a requirement in any of the reporting procedures that each line item of a procedure be in any way accounted for. senerally, the reporting procedures have strongly encouraged briefness in reports in the absence of a finding. This obviously will give rise to situations where the 766 percent completton data may not seem to agree with the reports. Further, the implementation of the 1981 prioritization scheme and the collateral decision to only do priority 1 and 2 level inspections sharply curtailed many of the RIV:RSk RTTaylor:gh 4ps /86 8708210026 B70919 ADOCK 0500 5
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T. F. Westerman procedural line items. The 766 date was adjusted in 1981 or 82 (by this writer) to reflect completion of a given procedure if all priority 1 and 2 level inspections had been already completed. As to the coment on unresolved and open items which should/could have been violations or deviations, this writer sharply disagrees with Mr. Phillips.
Both the NRC management and utility management paid a good deal of attention to unresolved items (open items did not exist until recently) and few unresolved items were written that did not meet the criteria for an unresolved item. Regional management was insistent that unresolved items be justified as such. Mr. Phillips is perhaps correct on the matter of corrective actions, however, 10 CFR 2.201 requires 1
the NRC to " concisely" state the violation which certainly encourages, and perhaps requires, the licensee to " concisely" respond. Mr. Phillips' interpretation of requirements is not consistent with the verbiage of 10 CFR 2.201 even though this writer does essentially agree with Mr. Phillips' interpretation. and Appended Charts:
It appears that Mr. Phillips has reclassified all reported findings into the various Appendix B criterion where he now believes they should have been since (a) unresolved items are not generally written in such a way as to denote a criterion that may have been violated and (b) essentially all violations were originally written against Criterion V since from at least 1975 to around 1982 the unofficial " enforcement policy" prohibited violations of criteria where
" measures were established" and were part of the SAR. Although unofficial, this policy was effectively implemented since all violations had to be concurred in by the Enforcement Coordinator. As to the broad assertion that a substantial number of unresolved items should have been violations, this writer sharply disagrees. As Mr. Phillips points out, a great deal of research would be required to support his assertion. This writer believes that the various inspectors of record with the facts of a particular situation fresh, were entirely capable of making the proper determination as opposed to someone trying to second guess them years after the fact.
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i With the exception of the coment in 1.b., the rest of the coments of j
paragraph 1 are not warranted.
Substantially none of the procedures referenced were part of the NRC program from 1974 to 1980 and by 1981, even those which were, were sharply curtailed by the priority scheme. As to 1.b.,
Mr. Phillips is apparently basing his view of the relative shortness of the mid-term inspection on some previous experience. The procedure (35200) existent in 1978 did not require anymore time than was put in at CPSES. The balance of the coments in this enclosure largely pertain to apparent 1
inconsistencies between 766 data and report discussion previously addressed.
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j 3-T. F. Westerman l
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Conclusion:==
It is apparent that Mr. Phillips' review of the inspection history relied on
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the inspection program baseline existent today without regard to the inspection program evolutions since CPSES started into construction in late 1974. Any attempt to use the 766 data without a knowledge of CPSES and program history, as Mr. Phillips apparently did, can only lead to erroneous conclusions.
It is also apparent thet he almost completely disregarded any contribution by the resident inspectcr in his review. The resident and regional programs were essentially separate but complementary. The overall objective was to have a good picture of each construction activity and exercise judgements as to what and what did not need additional attention without any serious attempt to absolutely account for every line item of every procedure.
I have attached a page of MC2512 of September 17, 1981, that fully endorses this approach.
R. G. Taylor Reactor Inspector Attachment C
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t, NUCLEAR REGULATORY COMMISSION J'
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REGION IV
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"tii ni"JS'tE*Mv'i*iu'EEE AMUNGTON. TEXAS 75011 l.,. ~ '
In Reply Refer To:
Dockets: 50-445/85-07 50-446/85-05 a
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!.c Texas Utilities Electric Company ATTN:
M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 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
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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 inspection.
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 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 report.
1'l During this inspection, it was found that certain of your activities were in violation of NRC requirements.
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 letter.
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Texas Utilities Electric Company 2
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, i
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l Dorwin R. Hunter, Chief l
Reactor Project Branch 2 I
Enclosures:
1.
Appendix A - Notice of Violation 2.
Appendix B - NRC Intpection Report 50-445/85-07 3.
50-446/85-05 cc w/ enclosure:
Texas Utilities Electric Company Skyway Tower i
400 North Olive Street
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Lock Box 81
'1 Dallas, Texas 75201 Texas Utilities Electric Company ATTN:
J. W. Beck, Vice President Skyway Tower 400 North Olive street t.ock Box 81 Dallas, Texas 75201
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1 APPENDIX A NOTICE OF VIOLATION Texas Utilities Electric Company Docket: 50-445/85-07 Comanche Peak Steam Electric Station 50-446/55 05 C. IR-3 2L Units 1 and 2 Permit:
CPPR-W 1
I During an NRC inspection conducted on April 1 through June 21~, 1985, l
violations of NRC requirements were identified. The violations involved a l
failure to correct RTE Delta hardware problems; failure to use class "E" concrete (grout) as specified; failure to document unsatisfactory conditions during receipt inspection of the Hydrogen recombiners; failure to inspect concrete mixer blades quarterly; failure to document cement scales which were j
out of calibration; failure to translate design criteria for reactor vessel I
installation into specifications, procedures, and drawings; failure to maintain stated tolerances and to report the failure on a nonconformance report; failure to audit RPV specifications / procedures, installation, and as-built records; failure to properly identify a charging system spool piece; and failure to i
furnish or maintain records for reactor coolant systems materials.
In accor-
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dance with the " General Statement of Policy and Procedures for NRC Enforcement
. Actions "10 CFR Part 2, Appendix C (1985), the violations are listed below:
1.
Failure to Promptly Correct an Identified Problem with RTE - Delta Potential Transformer Tiltout Subassemblies
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10 CFR 50, Appendix B, Criterion XVI as implemented by the Texas Utilities Generating Company (TUGCO) Quality Assurance Plan, Section_
Revision
, requires that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficien-cies deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, a potential problem with RTE - Delta potential transformer tiltout subassemblies, which are used in the emergency diesel generator control panels, was identified to the applicant via a letter, dated June 15, 1983, from Transamerica Delaval Inc. This letter also provided instructions for correcting the potential problem.
However, the applicant did not perform the corrective action. The NRC initially reported this item as unresolved in NRC Inspection Report 445/84-40.
This is a Severity Level V Violation.
(Supplement II.E) (445/8507-01 446/8505-01),
f 2.
Commercial Grout Used in Lieu of Class "E" Concrete l
10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QA
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f Plan, Section 5.0, Revision
, requires that activities affecting quality /
l shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
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1 Contrary to the above, commercial non-shrink grout was used to grout the Unit I reactor coolant pump and steam generator supports in lieu of Class "E" concrete as specified in Section 6-6 of drawing 2323-51-0550, l
Revision 4.
The NRC initially reported this item as unresolved in NRC IR 445/84-16.
This is a Severity Level IV Violatinn.
(Supplement II.E) (445/8507-02).
j 3.
Hydrogen Recombiners - Out of Specification Voltage Recorded on Westinghouse Quality Release Document 1
10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QA l
Plan, Section 5.0., Revision 2, requires that activities affecting quality shall be prescribed by documented instructions, procedures or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Brown aad Root Procedure QI-Q' ?-7.2-8, " Receiving of Westinghouse Safety Related Equipment", Section 3.1.d.1, required a QC inspector to verify that the Westinghouse Quality Release (QR) document checklist items be filled out completely and accurately.
4 Contrary to the above, the voltage recorded on Westinghouse QR 41424 checklist, attachment 1, step 4.1, was.outside the specified tolerance, l
but the QR was accepted as satisfactory by quality control receipt inspection. The NRC initially reported this item as unresolved'in NRC IR No. 50-445/84-16.
This is a Severity Level V Violation Supplement II.E (445/8505-03).
i 4.
Failure to Provide Objective Evidence (Records) to Show that Concrete Central and Truck Mixer Blades were Inspected 10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QA Plan, Section 5.0, Revision 2, dated May 21, 1981, requires that activities affecting quality shall be prescribed by documented instruc-tions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accorda Me with these instructions, proce-dures, or drawings."
Brown & Root Procedure 35-1195-CCP-10, Revision 5, dated December 4, 1978, requires that central and truck mixer blades be checked quarterly to assure that mixer blade wear does not exceed a loss of 10% of original blade height.
Contrary to the above, on May 31, 1985, the NRC inspector determined that there was no objective evidence (records) that the mixing blades had been inspected quarterly since the trucks were placed in service in 1977.
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This is a Severity Level V Violation.
(Supplement II.E) (445/8507-04; 446/8505-02).
5.
Failure to Issue a Deficiency Report on Cement Scales That Were Out of Calibration 10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QA Plan, Section 5.0, Revision 2, dated May 21, 1981, and Section 15.0, I
Revision 4, dated July 31, 1984, requires that activitias affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Brown & Root Procedure CP-QAP-15.1, " Field Control of Nonconforming Iten!, " states that nonconforming conditios shall be documented in a Deficiency and Disposition Report (DDR).
Procedure CP-QCP-1.3, " Tool l
Equipment Calibration and Control," dated July 14, 1975, states that out-of-calibration equipment shall be identified on a DDR.
Contrary to the above, on May 31, 1985, the NRC inspector reviewed the l
calibration file for scale (MTE 779) used for weighing cement and found
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that difficulty was encountered with the water and cement scales during a l
l 1975 calibration of the backup plant scales; however, no DDR was issued I
to identify this condition and require disposition of the scale and concrete (if any) produced.
This is a Severity Level IV Violation. (Supplement II.E) (445/8507-06; 446/8505-04).
6.
Failure to Translate Desian Criteria Into Installation Specifications, Procedures, and Drawings; and Failure to Control Deviations from These Standards 10 CFR 50, Appendix B, Criterion III, as implemented by TUGC0 QA Plan, i
1 Section 3, Revision 3, dated July 31, 1984, require that measures shall be established to assure that applicable regulatory requirements-and the design basis, are correctly translated into' specifications, drawings, procedures, and instructions. These measures shall include provisions to
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assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.
i Contrary to the above, (a) Unit 2 reactor pressure vessel installation i
l design criteria recommended by the NSSS vendor, such as centering tolerances, levelness tolerances, and shoe to bracket clearances, were not included in installation specification, procedures, and drawings; and (b) the criteria were specified in Construction Operation Traveler ME-79-248-5500, but were not treated as design engineering criteria as evidenced by an undocumented change of shoe to bracket clearances.
This is a Severity Level IV Violation (Supplement II.E) (446/8505-06).
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Failure to Maintain Tolerances Stated and Failure to Report These Results on a Nonconformance Report 10 CFR 50, Appendix B, Criterion XV as implemented by the TUGC0 QA Plan,Section V, Revision 2 dated May 21, 1981, requires that measures shall be established to control materials, parts or components which do not conform to requirements; and nonconforming items shall be reviewed and accepted, f
rejected, repaired; or reworked in accordance with documented procedures.
Brown and Root Quality Assurance Manual, Section 16 dated March 27, f
1985, requires that unsatisfactory conditions identified on process control documents shall be identified on an Nonconforming Report.
1 Contrary to the above, clearances between the reactor vessel support brackets and support shoes were not within the tolerance stated in Construction Operation Traveler ME-79-248-55 and the condition was not i
l reported on a Nonconformance Report.
This is a Severity Level IV Violation.
(Supplement II.E) (446-8505-07) 8.
Failure to Audit RPV Specifications / Procedures, Installation and As-built Records 10 CFR 50, Appendix B, Criterion XVIII as implemented by the TUGC0 QA Plan, Section 18.0, Revision 2, dated July 31, 1984, requires that a cn_morehensive system of,f Cplanned and periodic audits be carried out to verify compliance with all 3
aspects of the quality assurance program.
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either Unit 2 reactor vessel installation specifications, placement j
8 procedures, actual hardware placement, or as-built records.
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This is a Severity Level IV Violation.
(Supplement II.E) (446/8505-08).
9.
Failure to Properly Identify a Spool Piece j
10 CFR 50, Appendix B, Criterion VIII, as implemented by the TUGC0 j
QA Plan, Section 8.0, Revision 0, dated July 1,1978, requires that measures be established for the identification and control of materials, parts, and components, including partial?y fabricated assemblies.
These j
measures shall assure that identification of the item is maintained by i
heat number, part number or other appropriate means.
Article NA3766.6 of ASME,Section III,1974 Edition, requires that the identification of material consist of marking the material with the applicable mtterial specification and grade of material, heat number or heat code of the material, and any additional marking required by this j
section to facilitate traceability of the reports of the results of all tests and examinations performed on the material.
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Contrary to the abovei spool piece 3Q1 (DWG No. BRP-CS-2-RB-76) had neither been marked with the material specification and grade, heat number or heat code of the material.
This is a Severity Level IV Violation (Supplement II.E) (446/8505-09) 10.
Failure to Furnish or Maintain Records for Material l
10 CFR 50, Appendix B, Criterion XVII, as implemented by the TIGC0 QA Plan, Section
, Revision requires that sufficient records be maintained to furnish evidence of activities affecting quality. The records shall include at least the following: operating logs, and the results of reviews, inspections, tests, audits, monitoring of work performance, and materials analyses.
Articles NB-2130 and NA3767.4 of ASME Section III, 1974 Edition requires a Certified Material Test Report for all pressure reta$ning material over 3/4 inch nominal size.
Contrary to the above, Certified Material Test Reports were not available for the 22 degree elbow, 10 inch 45 degree nozzle, and three thermowell bosses, which were a part of the loop 3 cold leg piping subassembly.
This is a Severity Level IV Violation (Supplement II.E) (446/8505-10) l Pursuant to the provisions of 10 CFR 2.201, Texas Utilities Electric Company is hereby required to submit to this office, within 30 days of the date of this Notice, a written statement or explanation in reply, including:
(1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown.
Dated:
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APPENDIX B J
U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report:
50-445/85-07 Permit: CPPR-126 50-44G/85-05 CPPR-127 Docket:
50-445; 50-446 Applicant:
Texas Utilities Electric Company (TUEC)
Skyway Tower j
400 North Olive Street Lock Box 81 Dallas, Texas 75201 l
Facility Name:
Comanche Peak Steam Electric Station (CPSES)
Units 1 and 2 Inspection At: Glen Rose, Texas Inspection Conducted: April 1, 1985, through June 21, 1985 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 j
Inspector Construction (April 1 - May 10,1985)
(pars. 1, 3, and 19)
D. E. Norman, Reactor Inspector Date (pars. 1, 12, 13, 14, and 19)
D. M. Hunnicutt, Section Chief Date Reactor Projects Branch 2
[ (pars. 1, 4, 5, 6, 7, and 19)
Approved:
j D. M. Hunnicutt, Section Chief, Date 1
Reactor Project Section B J
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l Inspection Summary Inspection Conducted April 1, 1985, through 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 l
four NRC inspectors.
1 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 l
grout was used to grout the Unit I 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 I
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 j
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 status. The inspection involved 355 inspector-hours onsite by four NRC inspectors.
Results: Within the sixteen areas inspected seven violations were identified (failure to provide objective evidence to show that concrete central and truck mixer blades were inspected, paragraph 8; failure to issue a deficiency report on cement scales that were out-of-calibration, paragraph 9c; failure to translate design criteria into specifications, procedures, and drawings, paragraph 12a.; failure to maintain RPV installation tolerances / document nonconformance, paragraph 12b.; failure to audit RPV specifications, procedures, installation, and as-built records, paragraph 12d.; failure to l
identify a spool piece, paragraph 14b.; and failure to maintain material records, paragraph 14c.
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l l DETAILS i
1 1.
Persons Contacted Applicant Perconnel 4
i M. McBay, Unit 2 Reactor Building Manager
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I B. Ward, Gen. Supt., Civil l
D. Chandler, QA/QC Civil Inspector j
W. Cromeans, QA/QC, TUGC0 Laboratory / Civil Supervisor I
- J. Merritt, Assistant Project General Manager i
- P. Halstead, Construction Site QA Manager
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.#C. Welch, QA Supervisor TUGC0 (Construction)
J. Walters, TUGC0 Mechanical Engineer K. Norman, TUGC0 Mechanical Engineer j
J. Hite, B&R Materials Engineer G. Purdy, B&R CPSES QA Manager
- Denotes those present* at May 10, 1985 exit interview.
- Denotes those present at June 10, 1985 exit interview.
The NRC inspectors also interviewed other applicant employees during this inspection period.
I 2.
Plant Status Unit 1 At the time of this inspection, construction of Unit 1 was 99 percent complete.
The fuel loading date for Unit 1 is pending the results of ongoing NRC reviews.
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 I fuel loading.
3.
Applicant Action on Previous NRC Inspection Findinas a.
(Closed) Unresolved Item 445/8440-02:
Potential Problem with Potential Transformer Tiltout Subassemblies.
By letter dated June 15, 1983, Transamerica Delaval notified the 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 letter also provided instructions for correcting the problem, i
However, the NRC inspector could not determine if the problem had t
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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 1 corrective action work activities were documented on startup I
work pemits Z-2912 (train A) and Z-2914-(train B). The Unit 2 work activities are being tracked as master data base (MDB) ites 3003-31.
The failure to promptly correct this identified problem is an apparent violation (445/8507-01; 446/8505-01).
b.
(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 2323-51-0550 was acceptable.
Design Change Authorization 21179 was issued to drawing 2323-51-0550 accepting the use of the commercial non-shrink grout. However, the failure to grout with Class "E" concrete as specified on the drawing at the time the work was accomplished is an apparent violation (445/8507-02).
c.
(Closed) Unresolved Item 445/8416-04:
Hydrogen Recombiners -
Out-of-S) edification Voltage Recorded on Westinghouse Quality Release )ocument Quality Release N-41424 was revised changing the specified voltage from 10+-2V to 12+-2V which put the questionable voltage.within specification limits. 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 (44,5/8505-03).
d.
(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 l
scope and objectives of the site surveillance program were unclear, lacking both purpose and direction.
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 compliptntary to the audit and inspection programs.
Since the TUEC audit group 1s not located on site, the TUEC surveil-lance program on site takes on added significance.
1 This item was discussed with the TUEC site QC manager who described 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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l TUEC has elected to defer responding to the violations pertaining to 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 report and other QA matters. 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 plan. 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 assure quality especially the overview of quality control effectiveness.
4.
Document Inspection of Site Dans The NRC inspector reviewed documents describing the inspection activities performed on the Squaw Creek Dam (SCD) and the safe shutdown impoundment (SSI) for impounding cooling water for the two units at CPSES. The purpose of the SCD is to impound a cooling lake for CPSES. A secondary reservoir (SSI) is formed by a channel connecting the SCD impoundment to the SSI.
Three documented inspections have been performed since 1980. The-inspections were:
a.
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 Resources.
Date of Inspection: June 10, 1980.'
l b.
Inspection on August 25, 1982, by registered professional engineers from Mason-Johnston & Associates, Inc., and Freese & Nichols, Inc.
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c.
Inspection on September 19, 1984, by a registered professional engineer from Nason-Johnston & Associates, Inc.-
The inspection activities consisted of visual inspections by inspection teams that included accompanying Texas Utilities Service, Inc. (TUSI),
and Texas Utilities Generating Company (TUGCD) representatives.
Photographs were taken as a part of the documentation. The data for the piezometer observations and the data for the surface reftrence monuments t
were reviewed by applicant personnel and Mason-Johnston engineers.
No items of significance were observed or reported by these inspection teams.
Slight erosion areas were observed and reported. 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 Associates. No signs of cracks, settlements, or horizontal movement at any location within the SCD or the SSI were 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 dams.
1 The state of Texas requires periodic inspections of these dams (principally the SCD) due to inhabited dwellings downstream".
The applicant has met these inspection requirements.
No violations or deviations were identified.
5.
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 j
walls to elevation 860 feet 0 inches. When the forms were removed.
l honeycombs and voids were observed by applicant personnel.
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 testing.
In August 1982, preparations were made to pour the concrete annulus around the reactor vessel. When the expanded metal forework was removed from the reactor side of the compartment walls, voids were observed and NCR C-82-01202 was prepared. DCA 20856 was prepared as a procedure to repair the void area.
DCA 20856 indicated that the voids were not extensive (a surface area of about 28 square feet by 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 cubic yard of concrete was used to complete the repairs as indicated on grout pour card 261.
The applicant's review and evaluation of the gird pattern and a comparison of SG compartments 2 and 3 to 1 and 4 inclicated 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 liner. The liner walls of SG compartments 1 and 4
j i
l !
were not tested at elevation 834 feet, but at elevation 836 feet which is above the area of the identified voids.
No testing was done on the liner side of the area of the voids below elevation 836 4
feet.. The program also included removal of 2 inch x 2 inch plugs from the stainless steel liner at locations where test indications raised questions concerning the concrete. The inspections of the
.j 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 area 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
,1 area. Air access holes were drilled in the stainless steel liner plates to assure that grouting would be accomplished in accordance with the procedure.
The procedure (DCA-20856) specifed that the grout was to be cured I
for 28 days or until the grout reached a compressive strength of 4000 j
psi.
Repairs to the liner plates were specified in DCA-20856 and G&H Procedure 2323-55-18.
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 permitted.
Documentation' review indicated that DCA-20856 was adhered.to and that no cutting or drilling occurred in prohibited locations.
1 t
No violations or deviations were identified.
I a
6.
Nondestructive Testina Observations of Liner Plates in Fuel Transfer q
l Canal l
The NRC inspector 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 t
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 20856.
No violations or deviations were identified.
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7.
Rebar Placement and Cadweld Splice Observations and Reccrds i
The NRC inpector reviewed the rebar placement and Cadwell Splice activities associated with the Unit 2 containment (reactor building) l closure.
l a.
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 machina 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 l
inspector and recorded as follows:
Nominal load Calibration Readina Error Error Remarks (1bs)
(lbs)
(lbs) i 0
0 0
0 0 machine on 4/2/85 l
100,000 99,750
+250
+0.25 200,]00 199,600
+400
+0.2 300,00 299,450
+550
+0.18 l
350,000 350,300
-300
-0.08 400,000 401,200
-1200
-0.03 500,000 501,350
-1350
-0.27 l
600,000 602,450
-2450
-0.40 l
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 Procedure 35-1195-IEI-37, Revision 3, dated March 11, 1982. The reference standards were identified as follows:
ID No.
Manufacturer Calibration Due Date RS-75 BLH Electronics January 27, 1987 RS-75.3 BLH Electronics January 27, 1987 b.
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 caduelder qualification.
EBD Q8, GBH Q1, GBH Q2, GBV Q1, BFD Q4, BF0 Q3, BFH Q4, GAH Q1, l
GAV Q1, and GBV Q2.
j i
Each of the above qualification cadweld splices was tensile 1
tested to 400,000 pounds (100,000 psi) and met the requirements stated in the procedure.
j l
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. (2) Production Tensile Testing The NRC inspector observed the tensile tester calibrations and.
the following production cadweld splices tensile testing on May 7, 1985: FXD 3P, FYD 4P, FYD 8P, FR0 87P, and FUD 6P.
Each of the above production cadweld splices was tested to 400,000 pounds (100,000 psi)and met the requirements stated in the procedure.
1
('3) Installation of Production Cadweld Splices The NRC inspector observed installation of rebar and cadweld splices at frequent intervals (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 2 closure was performed in the area identified as elevation 805' feet to elevation 875 feet and azimuth 300 degrees.to 335 i
degrees. The installation activities observed included rebar
'; pacing, 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 es-built drawings.
D cementation Reviewed (4) f The NRC inspector reviewed the following documentation for the rebar placement and cadwelding for the Unit 2 containment (reactor building) closure area:
Orawinas DCAs NCRS 2323-S-0785, Rev.7 22616, Rev. 1 C85-200294 2323-5-0786, Rev.9 22728 C85-200339, Rev.1 l
2323-S1-500, Rev.5 22737 C85-200355, Rev.1 l
2323-S1-506, Rev.5 22836 2323-S2-505, Rev.5 22878 (Sheets 1-7) i 2323-52-508, Rev.2 22772 2323-52-506, Rev.3 No violations or deviations were identified.
8.
Concrete Batch Plant Inspection' The NRC inspector used a nationally recognized checklist to inspect the concrete production facilities. This list included the specific characteristics for the following areas:
(1) material storage and handling of cement, aggregate, water and admixture, (2) batching equipment scales, weighing systems,-adeixture dispenser, and recorders, (3) central mixer (not applicable because it had been dismanteled), (4) ticketing system, and (5) delivery system.
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i The current batching is a manual operation since almost all concrete has The central mixer was dismanteled and removed from site two been placed.
or three years ago when concrete placement was virtually completed Presently, the backup batch plant (which was a backup system for the central mixer) is in operation to complete the remaining concrete placements. This batch plant is in good condition and complied with the subject checklist except for one area The NRC inspector inspected the inside of one of three trucks used for mi'xing 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 The subject to wear and should be checked at a reasonable frequency.
l Brown & Rost (B&R) representative responsible for checking the blades in i
accordance with 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
I such checks dating back to 1977 when they were initially checked, i
i 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 J
would have probably indicated if there was a problem with the mixing blades.
Strength and uniformity tests have consistently been within the acceptable i
range indicating that concrete production was acceptable even though mixing blade inspection was not documented.
Otherwise, the condition of the-inside of the truck was satisfactory as the drum and charging / discharging were clean. The water gage and drum counter were in good condition.
This failure to follow procedures is a violation of 10 CFR 50, Appendix B, Criterion V.
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).
l No other violations or deviations were identified.
9.
Calibration Laboratory for Batch Plant j
1 The NRC inspector obtained batch plant scale numbers from tags which indicated that the scales had been calibrated and were within the calibration frequency. 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 activated. The d
records were adequate except as follows:
Scales MTE.766 records do not clearly differentiate between the a.
required accuracy of the scale and the digital readout.
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, b.
Scales MTE 779 and 780 records show various accuracy ranges for the same scale; i.e., MTE 779 (SN749687) records the following: ' report i
dated January 1976 gives 1%; 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 proper.
I l
c.
Records for scales MTE 779 records contained'8&R memo IM-1108 dated j
l July 16, 1975, which described a nonconforming condition. This condition affected the water and cement scales causing a 24-48 pound deviation during the calibration test. The meno stated that the i
i condition was corrected and the scales were then calibrated; however, no deficiency report was yritten 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 l
Nonconforming Items," dated July 1A, 1975. As a result there is'no evidence that corrective action included an evaluation to determine if concrete production was adversely affected.
I l
This failure to assure that a nonconforming condition was evaluated j
is a violation of Criterion XV of 10 CFR Part 50, Appendix B.
J (445/8507-06; 446/8505-04).
10.
Concrete Laboratory Testing l
l 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 procedure.
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 frequency. Test personnel were knowledgeable of test requirements and equipment.
The NRC inspector witnessed field tests performed by laboratory personnel as follows:
Date Truck No. Mix No. Ticket No.
Air Content (%) Slump Temperature
.U.n.:,1
,,F, 6/3/85 RT-41 925 64013 Req 8.2-10.3 NA 70 max Mea 8.7-9.1 NA 57 6/3/85 RT-35 128 64014 Req 5.0-7.0 5 max 70 max Mea 6.6 6.25*
57
- Truck was rejected by quality control but was later accepted when second slump reading came into required range.
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The following laboratory equipment was checked and found to be within calibration:
Forney Compression Tester, MTE 3033: Temperature Recorder.
)
MTE 3013 and 3014; Unit Volume Scale, MTE 1053; Pressure Meters MTE 30008, 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 sold MTE 1111.
The following test records for placement number 201-5805-034 were reviewed: (1) concrete placement inspection, (2) concrete placement summary anM, (3) unit weight of fresh concrete.
No violations or deviations were identified.
11.
Inspection of Level C and 0 Storace The NRC inspector inspected all laydown areas where piping, electrical conduit, cable, and structural reinforcing steel were stored.
These materials were neatly stored outside on cribbing in well drained areas i
which allowed air circulation and' avoided trapping water. This met the j
Level "0" storage requirements of ANSI N45.2.2.
1Property "ANSI code" (as page type) with input value "ANSI N45.2.2.</br></br>1" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. The electrical warehouse contained miscellaneous-electrical hardware.
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 them.
The first problem noted was that a box of nuclear grade cement was marked
" shelf life out of date" but it had no hold tag.
The box was.
subsequently tagged with Nonconformance Report (NCR) E85-200453 after 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 tag. 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 leak was in an area that did not expose hardware to moisture.
The roof is currently being repaired.
The millwrir,ht 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 requirements.
No violations or deviations were identified.
- 12. 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 internals by examining the completed installation and inspection records.
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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 (WNS0) " 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 recommendations:
o WNSD drawing 1210E59 " Standard - Loop Plant RV Support Hardware Details and Assembly" WNSD drawing 1457F27 " Comanche Peak SES RCS Equipment Supports o
- Reactor Vessel Supports" o
CE drawing 10773-171-004 " General Arrangement Elevation" o
CE drawing 10773-171-005 " General Arrangement Plan" I
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 shoes. The lack of engineering documentation did not provide full control of the.
action and would allow changes to installation criteria important to l
safety to be made'without complying with established change procedures.
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 RPV.
Requirements recommended by WNSD
)
procedures were implemented in the traveler. Worksheets attached to I
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 j
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 38 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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t 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 and 26 0-Rings o
Deviation notices and corrective action statements The NRC inspector reviewed the following completed travelers for i
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 ME-81-2145-5500,'"Retorque UI Column Extension" o
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" c.
V_isual Inspection At this time, visual inspection of the internals by the NRC inspector was not possible, and inspection was limited or. 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 area.
d.
Records of QA Audits or Surveillance j
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 as-built records.
Failure to perform audits or surveillance of RPV specifications, procedures and installation is a violation of 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 paragraphs.
i 13.
Reactor Vessel Disorientation On February 20, 1979, T.he 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 degrees.
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
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i I design change was implemented for the reactor vessel, but not for the pedestal support locations.
The problem was not considered by the
?
applicant to be reportable under provisions'of 10 CFR Part 50.55(e) since 4
the error could not have gone undetected.
4 l
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, t
y l
Maryland, the applicant presented the proposed redesign and rework
'l procedures for relocating the pedestal supports. No unresolved safety 7')
concerns with the repair were identified at the meeting.
l l
During this inspection the NRC inspector reviewed.various documentation
,l relative to the disorientation problem, including design changes and the construction traveler which implemented the repair.
The following documents were reviewed:
]
o NRC Inspection Report 50-446/79-03 o
NRC Inspection Report 50-446/79-07 I
o NRC Inspection Report 50-446/79-13 I
o TUSI Conference Memo, dated March 1, 1979, H. C. Schmidt to S. Burvell-(NRC Licensing PM) l
,y 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 j
Structure for Placement of the RPV Support' Shoes o
DCA 4122, dated March 22,1979, ' Sui > ject:. Replacement of Rebar for j
l RPV Supports o
Construction Traveler CE79-018-5505, dated March 14, 1979,
Subject:
-l 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' identified.
14.
Reactor Coolant Pressure Boundary (RCf'B) Systems The inspection was performed to verify:
the applicants system for preparing, reviewing, and maintaining records for the. RCPB piping and components; that selected records reflected compliance with NRC requirements and SAR commitments for manufacture,Ltest and installation I
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l i of items; and that as-built hardware was adequately marked and traceable to records.
The following items were randomly selected and inspected:
a.
Pressurizer Safety Valve This item was inspected to the commitment stated in FSAR, Table 5.} 1 which ipcludes ASME Section III, 1971
)
edition through winter 1972 addenda. Valve S/N N56964-00-007, which is installed in the B position was inspected.
The following records were reviewed; QA Receiving Inspection Report No. 21211 o
o Code Data Report Form NV-1 Valve Body'CMTR o
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; records.
j l
b.
CVCS Scool Piece 3Q1 - Requirements for this item are stated in l
ASME,Section III,1974 edition through summer 1974 addenda, which 1
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 I
I welds number 1 and 3 (ref. BRP-CS-2-RB-076).
The following records l
were reviewed:
o B&R Code Data Report o
Field Weld Data Card 4
o NDE Reports I
QA Receiving Reports (for bulk order) o o
Certified Material Test Report (CMTR)
The installed spool piece was inspected for weld quality and to verify that marking and traceability requiremenic'had been met. The item had been u d kod with the spool piece number (3QI) and the B&R l
drawing number, however, marking of the material specification l
g number and type, heat code, or other means of traceability could
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not be found, 'In respect to material requiring WCMTR, (nominal pipe size greater than 3/4 inch) NA-3766'Pequires marking with the l
applicable specification and grade of materiar and hent number or i
heat code. 'When material is didded, the identification marking is
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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, c.
Section III, 1974 edition through summer 1974 addenda, which is the commitment from the FSAR, Table 5.2-1.
This piping subassembly
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. 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) dW 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 l
(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 l
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 l
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 I
(446/8505-11).
(3) The cold leg NPP-1 Form stated that no hydrostatic (Hydro) test had been performed.
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 test.
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 NS-6221(a) shall be acceptable as a test for piping subassemblies.
NB-6221(a) states that completed components shall be subjected to a hydrostatic test prior to installation in the system.
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, Paragraph (b) of NB-6221 requires pressure testing of all 4
'Tressure retaircing cocoonents that are within the boundary
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piot'ected by overpressure protection devices and paragraph (c)
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permits substitution of the system test for component test.
It-H is not evident th 6 the system test substitution was permitted for pipe basseeMy since NA-1200 makes a distinction in the definitio,w;between components and piping subassemblies.
I NB-6115 n
states that pressure testing of components shall be performed
-pr1or to initial operat.*sn of a system tnd that the Data Report ',
Fern rihall tob be completed nor signed by the code inspector.and 1
3 the' component.shall not be stamped until the component manufac-
?,
turer has conducted the hydrostatic pressure test. Additionally, NA-8231 prohibits the bplication of a Code stamp prior to hydrostatictestr*.Mrdlessofwhetherthetestwasperformed prior to or aften, installation of the item. The NPP-1 Form for.
the cold leg had bein completed and signed by> td.e maNfECturer and Authorized Nuclear Inspector (ANI). f.o A51E CodeJtemp har' % ( is t
4,,4 also been'appd ed to the ften. The NRC inspect v observed that a j,
hydrostatic te:t had not been performed and we.s noted on the A
NPP-1 ferm.:
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The ab ve items are unresolved pending clarification of code 4
requirements by NRC headquarters (446/0505)-12).
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l hs) Since the cold leg pipe subassembly had not been pressure
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tested prior to installation, the NRC inspectud reviewed the procedures and hydro test data applicabie to Unit 1, since
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Unit 2 hydro hacoct been completed.. kequiremerits for the-tests u
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wdr'e' presented in Procedures CP-QAP-12.2, "In nection Procedure..
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'I aqd Acceptance Crite"f a for ASME Prassure 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 ts tm used in the test, be prepared with the test boundary and the A-additional following data shown:
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Base metal defects in which filler material has been added, and the depth of the base metal defect exceeds 3/8 m inch or 10% of the actual thickness, whichever is less.
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' Untested vendor performed piping circumferential welds.
o Approximate location and material identification and o
description for permanent pressure t,oundary attachment with applicable apport number referenced, l
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Weld history, whkh shall reflect wnld removal and/or waid o
repair.
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cold leg was reviewed for compliance with the above requirements. Drawing No. BRP-RC-1-520-001 inad been used to annotate the test boundary.
A handwritten statement on the drawing indicated:
"No major base metal repairs could be l
located" and "No hangers with weld attachments could be located." Welds performed by the pipe subassembly vindor, including the 22 degree circumferential weld and the penetration fittings had not been identified. The following l
items were unresolved regarding the adequacy of the hydro test:
o Was the determination of no major base metal repairs based l
on a visual inspection or on a review of vendor and site j
inspection and repair records?
l 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?
l The above issues will remain unresolved pending further I
evaluation by the applicant (445/8507-07; 446/8505-13).
d.
Personnel Qualifications - Personnel who had performed selected tasks were identified during inspection of installation records.
Training and experience records for the personnel were revi ued to verify j
that employee qualifications and maintenance of records were current 1
and met requirements. Names or codes for five welders and two NDE l
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 inspection.
I 15.
Special Plant Tours (Unit 1 and Unit 2) l On May 23, 1985, the NRC inspector conducted a tour of selected areas of Unit I and Unit 2.
The group consisted of one NRC inspector, two NRC Technical Review Team (TRT) representatives, two allegers, and several TUEC representatives. The TUEC representatives tagged each area where a i
deficiency was alleged. With the alleger's consent, a tape recorder was l
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 analyzing this information and will decide what action, if any, thould be l
taken.
During this tour the NRC inspector independently identified a p
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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stated that areas on tne wall were reserved for pipe hangers GHH-S1-1-SB-038-006 and R1(?)1-087-X11.
These stickers were dated 1980.
It was ??? evident whether hangers missing or none were needed in these i
locations and the reserve tags were not removed. TUEC representatives were unable to answer the question immediately. This item is unresolved pending further review during a subsequent inspection.
(445/8507-07, 446/8505-05).
1 No violations or deviations were identified.
16.
Routine Plant Tours (Units 1 and 2) i 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 building.
l 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' activities.
No violations or deviations were identified.
t l
17.
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 l
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 submitted.
No violations or deviations were identified.
18.
Review of Violation and Unresolved Item Status The NRC inspector reviewed all violations and unresolved items reported 3
to date to assure that NRC and TUEC status logs were complete and up to 1
date. Two hundred nineteen items were reviewed.
In addition, a trend 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 B.
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 j
Material, Equipment and Services; IX.
Control of Special Processes; X.
l l
Inspection; XI. Test control; XIII. Handling Storage and Shipping; XVII.
QA Records; and XVIII Audits.
The most significant trends were Criterion III, V, VII, IX, X, and XVIII. Also, a number of violations occurred with respect to 10 CFR 50.55(e) items.
These findings mainly pertained to Unit 1 and related closely to trends identified by the NRC Technical Review Team TM. - These trends will be considered during followup on TRT findings.
Also, Unit 2 inspection emphasis will consider these trends during future inspections.
No violations or deviations were identified.
i
... 1 19.
Exit Interviews The NRC inspectors met with members of the TUEC staff (dencted 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 l
ATTN:
M. D. Spence, President, TUGC0
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j Skyway Tower 400 North Olive Street
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Lock Box 81 Dallas, Texas 75201 Gentlemen:
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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 inspection.
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 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 report.
l During this inspection, it was found that certain of your activities were in violation of NRC requirements.
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 letter.
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j RPB2/B C/RPB2B C/RPB2 NRR HSPhillips/dc 0Hunnicutt 0 Hunter VNoonan
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Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, i
Dorwin R. Hunter.. Chief Reactor Project Branch 2 i
Enclosures:
1.
Appendix A - Notice of Violation 2.
Appendix B - NRC Inspection Report 50-445/85-07 50-446/85-05
,\\
cc w/ enclosure:
Texas Utilities Electric Company j
Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Texas Utilities Electric Company ATTN:
J. W. Beck, Vice President Skyway Tower 400 North Olive street l
Lock Box 81 l
Dallas, Texas 75201 bec distrib. by RIV:
J TEXAS STATE DEPARTMENT OF HEALTH bec: to DMB(IE01) bec distrib by RIV:
"RPB1
- Resident Inspector OPS
- RPB2
- Resident Inspector CONS R. Martin, RA
- D. Hunnicutt, Chief, RPB2/8 R. Denise, D/DRSS V. Noonan, NRR C. Wisner, PA0 S. Treby, ELD MIS SYSTEM RIV File i
Juanita Ellis Renea Hicks
- D. Weiss, LFMB (AR2015)
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In Reply Refer To:
Dockets: 50-445/B5-07 50-446/85-05 l
Texas Utilities Electric Company g'
g ATTN:
M. D. Spence, President, TUGC0 N
Skyway Tower h5L) 400 North Olive Street
/
Lock Box 81 Dallas, Texas 75201 Gentlemen:
This refers to the inspection conducted undar the Resident Ins mction Program by Mr. H. S. Phillips and others during the period April 1,19%, through June 214 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 inspection.
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 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 report.
I During this inspection, it was found that certain of your activities were in violation of NRC requirements. 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. litter.
C/RPB21%@tt C/RPS2 NRR RPB2/8 HSPhillips/dc DHunnicu DHunter YNoonan F/l /85
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Texas Utilities Electric Company 2
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely.
l l
R. P. Denise Director Division of Reactor Safety and Projects
Enclosures:
1.
Appendix A - Notice of Violation 2.
Appendix B - NRC Inspection Report 50-445/85-07 50-446/85-05 J
cc w/ enclosure:
Texas Utilities Electric Company Skyway Tower i
400 North Olive Street I
Lock Box 81 Dallas, Texas 75201 Texas Utilities Electric Company ATTN:
J. W. Beck, Vice President Skyway Tower 400 North Olive street Lock Box 81 Dallas, Texas 75201 bec distrib. by RIV:
TEXAS STATE DEPARTMENT OF HEALTH bec: to DM8(IE01) bec distrib by RIt:
- RPB1
- Resident Inspector OPS
- RPB2
- Resident Inspector CONS i
R. Martin, RA
- D. Hunnicutt, Chief, RPS2/B R. Denise, D/DRSS V. Noonan, NRR C. Wisner, PAD S. Treby, ELD MIS SYSTEM RIV File Juanita Ellis Renea Hicks
- D. Weiss, LFM8 (AR2015)
- w/766
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APPENDIX _A.
NOTICE OF VIOLATION Texas Utilities Electric Company Docket: 50-445/85-07 Comanche Peak Steam Electric Station 50-446/85-05 Units 1 and 2 Permit: CPPR-126 CPPR-127 During an NRC inspection conducted on April 1 through June 21, 1985, violations of NRC requirements were identified. The violations involved a 4
failure to correct RTE Delta hardware problems; failure to use class "E" l
concrete (grout) as specified; failure to document unsatisfactory conditions i
during receipt inspection of the Hydrogen recombiners; failure to inspect concrete mixer blades quarterly; failure to document cement scales which were out of calibration; failure to translate design criteria for reactor vessel installation into specifications, procedures, and drawings; and failure to maintain stated tolerances and to report the failure on a nonconformance report.
In accordance with the " General Statement of Policy and Procedures for NRC Enforcement Actions "10 CFR Part 2, Appendix C (1985), the violations are listed below:
1.
Failure to.Promptly Correct an Identified Problem with RTE - Delta Potential Transformer T11 tout Subassemblies 10 CFR 50, Appendix 8 Criterion XVI as implemented by Texas Utilities Generating Company (TUGCO) Quality Assurance Plan (QAP), Section 16.0, Revision 0, requires that seasures shall de established to assure that conditions adverse to quality, such as failures, malfunctions, deficien-cies deviations, defective material cnd equipment, and nonconformances are
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promptly identified and corrected.
Contrary to the above, a potential problem with RTE - Delta potential
/
transformer tiltout subassemblies, which are used in the emergency diesel generator control panels, was identified to the applicant via a letter,
- dated June 15, 1983, from Transamerica Delaval Inc. This letter also provided instructions for correcting the potential problem. However, the applicant did not take the corrective action. The WRC initially reported this ites as unresolved in NRC Inspection Report 445/84-40.
This is a Severity Level V Violation. (Supplement II.E) (445/8507-01 446/8505-01).
2.
Commercial Grout Used in Lieu of Class *E" Concrete 10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QAP, I
Section 5.0, Revision 2 requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accom dance with these instructions, procedures, or drawings. plished in accor-0
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Contrary to the above, commercial non-shrink grout was used to grout the k^
Unit i reactor coolant pump and steam generator supports in lieu of Class gp
- E' concrete as specified in Section 6-6 of drawing 2323-S1-0550, Revision 4.
M l, This is a Severity Level IV Violation.
(Supplement II.E) (445/8507-02).
3.
Hydrogen Recombiners '- Out of Specification Voltage Recorded on Westinghouse Quality Release Document 10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QAP,
)
Section 5.0., Revision 2, requires that activities affecting quality shall be prescribed by documented instructions, procedures or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Brown and Root Procedure QI-QAP-7.2-8, " Receiving of Westinghouse Safety Related Equipment", Section 3.1.d.1, required a QC inspector to verify that the Westinghouse Quality Release (QR) document checklist items be filled out completely and accurately.
Contrary to the above, the voltage recorded on Westinghouse QR 41424 checklist, attachment 1 step 4.1, was outside the specified tolerance, but the QC receipt inspector accepted QR as satisfactory.
This is a Severity Level V Violation (Supplement II.E) (445/8507-03).
4.
Failure to Provide Objective Evidence (Records) to Show that Concrete central and Truck Mixer Blades were Insoected 10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QAP, Section 5.0, Revision 2, dated May 21, 1981, requires that activities affecting quality shall be prescribed by documented instructions,
'j f procedures, or drawings, of a type appropriate to the circumstances and dy shall be accomplished in accordance with these instructions, procedures,
\\[j or drawings.
4 Brown & Root Procedure 35-1195-CCP-10 Revision 5, dated December 4, 1 1978, requires that central and truck mixer blades be checked quarterly to assure that mixer blade wear does not exceed a loss of 105 of original e
blade height.
Contrary to the above, on May 31, 1985, the NRC inspector determined that there was no objective evidence (records) that the mixing blades had been inspected quarterly since the trucks were placed in service in 1977.
This is a Severity Level V Violation.
(Supplement II.E) (445/8507-04; 1
446/8505-02).
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5.
Failure to Issue a ()eficiency Report on Cement Scales That Were Out of Calibration 10 CFR Part 50, Appendix B, Criterion V, as implemented by the TUGC0 QAP, Section 5.0, Revision 2, dated May 21, 1981, and Section 15.0, Revision 4,
, dated July 31, 1984, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type
(
appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
/
Brown & Root Procedure CP-QAP-15.1, 'F.feld Control of Nonconforming item
- states that nonconforming conditions shall be documented in a Deficiency and Disposition Report (DDR). procedure CP-QCP-1.3, ' Tool l
Equipment Calibration and Control," ' dated July 14, 1975, states that i
out-of-calibration equipment shall be identified on a DDR.
W41*
Contrary to the above, on May 31, 1985, the NRC inspector reviewed the calibration file for scale (NTE 779) used for weighing cement and found QM, 4
that a 24-48 pound deviation from the required accuracy was encountered with the water and cement scales during a 1975 calibration of the backup s
() require disposition of the scaleplant scales, however, no DDR was issued to I
and concrete (if any) produced.
This is a Severity Level IV Violation. (Supplement II.E) (445/8507-06; 446/8505-04).
6.
Failure to Translate Design Criteria Into Installation Specifications.
Procedures, and Drawings; and Failure to Control Deviations from These standards 10 CFR 50, Appendix 8. Criterion III, as implemented by TUGC0 QAP, Section 3, Revision 3, dated July 31, 1984, requires that measures shall be established to assure that applicable regulatory requirements and the design basis, are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.
Contrary to the above, Unit 2 reactor pressure ve:sel installation criteria such as centerin tolerances, levelness tolernaces, and shoe to bracket clearances were:g(a) specified on Construction Operation Traveler ME-79-248-5500 based on NS$$ vendor recommended criteria which had not been translated into CPSES specifications, drawings, procedures, or instructions; and (b) changed on the traveler without appropriately documenting such engineering changes.
This is a Severity Level IV Violation (supplement II.E). (446/8505-05).
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Failure to Maintain Tolerances Required and Failure to Report Tolerance Deviations on a Nonconformance Report j
i 10 CFR 50, Appendix B, Criterion XV as im>1emented by the TUGC0 QAP, Sectior. 15.0, Revision 2 dated May 21, 1931, requires that measures shall be I
established to control materials, parts or components which do not conform to requirements; and nonconforming items shall be reviewed and accepted, rejected, repaired; or reworked in accordance with documented procedures.
Brown and Root Quality Assurance Manual Section 16. dated March 27, 1985, requires that unsatisfactory conditions identified on process control documents shall be identified on a nonconforming report.
j Contrary to the above, clearances between the reactor vessel support brackets and support shoes were not within the original or revised i
permissible tolerance stated in Construction Operation Traveler l
ME-79-248-55 and the deviation was not reported on a nonconformance i
report.
This is a Severity Level IV Violation. (Supplement II' E) (446-8505-06)
Pursuant to the provisions of ~10 CFR 2.201, Texas Utilities Electric Company is hereby required to submit to this office, within 30 days of the date of this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when j
full compliance will be achieved. Consideration may be given to extending your l
response time for good cause shown.
l Dated:
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1 APPENDIX B
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U. S. NUCLEAR REGULATORY C00941SSION REGION IV i
NRC Inspection Report: 50-445/85-07 Pemit: CPPR-126 50-446/85-05 CPPR-127 Docket: 50-445; 50-446 Applicant: Texas Utilities Electric Company (TUEC)
Skyway Tower 400 North Olive Street Lock Box 81
)
t Dallas, Texas 75201 4
Facility Name: Comanche Peak Steam Electric Station (CPSES)
Units 1 and 2
)
Inspection At: Glen Rose, Texas Inspection Conducted: April 1,1985, through June 21, 1985 Inspectors:
J/
/0 !ff H. 5. Phillips, Senior Resident Date l
Reactor Inspector Construction l
(pars. 1, 2, 3, 8, 9, 10, 11,.15, 16, i
17, 18, and 19)
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J. E. Cummins, Senior Resident Reactor Uati
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Inspector Construction (April 1 - May 10,1985)
(pars. 1, 3, and 19)
Y/
/df//W D. E. Norman, Reactor Inspector Date (pars. 1, 12, 13, 14, and 19) dk A=:~*-
W1hr D. M. Kunnicutt Section chief pate Reactor Projects Branch 2 (pars.1, 4, 5, 6, 7, and 19) w w ~s o / V o M S f!-.. -.-. - -
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D. M. Hunnicutt, Section Chief.
Date Reactor Project Section B Inspection Stannary Inspection Conducted April 1.1985, through June 21,1985(Report 50-445/85-07)
I Areas Inspected: Routine, announced and unannounced inspections of Unit 1 I
4 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 pnsite by, four NRC inspectors.
-7
_,m gpl Results: Within the areas inspected, tions were identified: fail-N ure to promptly correct an identified problem with RTE - Delta Potential Transformer Tiltout Subassemblies, paragraph 3.a.; commercial non-shrink grout was used to grout the Unit I reactor c.colant 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 but QC receipt inspector accepted, paragraph 3.c; failure to provide objective evidence to show that central and truck mixer blades were inspected, paragraph 8; and failure to issue a deficiency report on cement scales that were out-of-calibra-tion, paragraph 9.c.
Inspection Sumary 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 butiding, 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 status. The inspection involved 335 inspector-hours onsite by four NRC inspectors.
Results: Within the sixteen areas inspected five violations were identified:
failure to correct RTE-Delta transfomer problem, paragraph 3.a; failure to provide objective evidence to show that concrete central and truck mixer blades were insm:ted, paragraph 8; failure to issue a deficiency report on cement scales that were out-of-calibration, paragraph Sc; failure to translate design criteria into specifications, procedures, and drawings, paragraph 12a.; and failure to maintain RPV installation tolerances / document dev'ations in a nonconformance report, paragraph 12b.
l DETAILS 1.
Persons Contacted Applicant' Personnel M. McBay, Unit 2 Reactor Buf1 ding Manager B. Ward, General Superintendent, Civil D. Chandler. QA/QC Civil Ins actor W. Cromeans, QA/QC, TUGC0 La%ratory/ Civil Supervisor l
1
- fJ. Merritt, Assistant Project General Manager
J. Walters, TUGC0 Mechanical Engineer K. Norman, TUGC0 Mechanical Engineer J. Hite, 88R Materials Engineer G. Purdy, 88R CPSES QA Manager
- Denotes those present at May 10, 1985 exit interview.
fDenotes those present at June 10, 1985 exit interview.
The NRC inspectors also interviewed other applicant employees during this inspection period.
2.
Plant Status l
Unit 1,
(
At the time of this inspection, construction of Unit I was 99 percent l
complete. The fuel loading date for Unit 1 is pending the results of
(
ongoing NRC reviews.
Unit 2 At the time of this inspection, construction of Unit 2 was approximately l
74 percent complete. Fuel loading is scheduled'for approximately 18 months after Unit 1 fuel loading.
3.
Applicant Action on Previous WRC Inspection Findinos a.
(Closed) Unresolved Itse 445/8440-02: Potential problem with Potential Transformer T11 tout 5 subassemblies.
By letter dated June 15, 1983, Transamerica Delaval notified the applicant of an RTE - Delta 10 CFR Part 21 report to the NRC l
reporting a potential problem with.the primary disconnect clips of
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the potential transformer tiltout assembly used in the emergency diesel generator control panels at CPSES. The Transamerica Delaval letter also provided instructions for correcting the problem.
However, the NRC inspector could not determine if the problem ~had been corrected at CPSES and made this an unresolved ites. The applicant determined that the problem had not been corrected and l
subsequently performed the recommended corrective action. The Unit 1 corrective action work activities were documented on startup-l work permits Z-2912 (train A) and 2-2914 (train B). The Unit 2 work activities are being tracked as master data base (28) item 3003-31.
The failure to promptly correct this identified problem is an apparent violation (445/8507-01; 446/8505-01).
1 b.
(Closed) Unresolved Item 445/8416-03: Commercial GroJt Used in Lieu I
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 2323-51-0550 was acceptable. Design Change Authorization 21179 was issued to drawing 2323-S1-0550 accepting the use of the commercial non-shrink grout. However, the failure to grout with Class "E" concrete as specified on the drawing at the time the work was accomplished is an apparent violation (445/8507-02).
c.
(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+-2Y which put the questionable voltage within specification limits. 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 (445/8507-03).
c.
(0 pen) Unresolved Item 445/8432-06: 446/8411-06: Lobbin Report Described Site surveillance Program Meaknesses 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 direction.
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 a'so observed, as.did the Lobbin l
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.s-Report, that the surveillance program lacks both purpose and direction to be effective and complimentary to the audit and inspection programs.
Since the TUEC audit group is not located on site, the TUEC surveil-lance program on site takes on added significance.
This item was discussed with the TUEC site QC manager who described a reorganized site surveillance function and changes that have 1
occurred. New procedures which describe this organization's duties and responsibilities are forthcoming.
TUEC has elected to defer responding to the violations pertaining to the audit function in NRC Inspection ieport 445/84-32; 446/84-11, but rather to have the Comanche Peak Response Team (CPRT) respond to this report and other QA matters. The surveillance issue is closely tied to the audit deficiencies in NRC Inspection Report No. 445/84-32; 446/84-11. This ites will remain open pending the review and imple-mentation of the CPRT action plan. 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 assure quality, especially the overview of quality control effectiveness.
4.
Document Inspection of Site Dams The NRC inspector reviewed documents describing the inspection activities performed on the Squaw Creek Dam (SCD) and the safe shutdown impoundment 1551) for impounding cooling water for the two units at CPSES. The purpose of the SCD is to impound a cooling lake for CPSES. A secondary reservoir (SSI) is formed by a channel connecting the SCD impoundment to the SSI.
l Three documented inspections have been performed since 1980. The inspections were:
a.
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 Resources.
Date of Inspection: June 10,1980.
b.
Inspection on August 25, 1982, by registered professional engineers from Mason-Johnston & Associates, Inc.,, and Freese & Nichols Inc.
c.
Inspection on September 19, 1984, by a registered professional engineer from Mason-Johnston & Associates, Inc.
j 1
The inspection activities consisted of visual inspections by inspection teams that included accompanying Texas Utilities Service, Inc. (TUSI),
and Texas Utilities Generating Company (TUGC0) representatives.
1 Photographs were taken as a part of the documentation. The data for the I
l i
l l
i...
piezometer observations and the data for the surface reference monuments were reviewed by applicant personnel and Mason-Johnston engineers.
No items of significance were observed or reported by these inspection teams. Slight erosion areas were observed and reported. A cracked area on the service spillway upstream right bridge seat was observed by the inspection teams and continued. monitoring of this area was.recomended by Mason-Johnston and Associates. No signs of cracks, settlements, or horizontal movement at any location within the SCD or the SSI were reported.
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 dams.
The state of Texas requires periodic inspections of these dans (principally the SCD) due to inhabited dwellings downstream. The applicant has met these inspection requirements.
No violations or deviations were identified.
5.
Voids Behind the Stainless Steel Cavity Liner in Unit 2 Reactor Building The NRG inspector reviewed applicant records, including NGR G-52-01Z0Z; 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-SS-18. The review of 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 personnel. The.
appifcant'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-testing.
~-
! l j
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 WCR C-82-01202 was prepared. DCA 20856 was prepared as a procedure to repair the void area. DCA 20856 indicated that the voids were not extensive (a surface area of about 28 square feet by 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 cubic yard of concrete was used to complete the repairs as indicated ~
on grout pour card 261.
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 extertded down to elevation 834 feet, which is the floor elevation of the liner. The liner walls of SG compartments I and 4 were not tested at elevation 834 feet, but at elevation 836. feet which is above the area of the identified voids. No testing was done on the liner side of the area of the voids below elevation 836 feet. The program also included removal of 2 inch x 2 inch plugs 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 j
areas (voids).
The applicant removed stainless steel liner plates from three areas (one area 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 (1 1/4) diameter probe holes and grout access holes were drilled in the liner plates to detemine the extent of and to assure full definition of the void area. Air access holes were drilled in the stainless steel liner plates to assure that grouting would be accomplished in accordance with the procedure.
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-SS-18.
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 permitted. Documentation review indicated that DCA-20856 was adhered to and that no cutting or drilling occurred in prohibited locations.
No violations or deviations were identified.
r.
6.
Nondestructive Testing Observations of Liner plates in Fuel Transfer Canal The NRC inspector observed portions of non-Q liquid penetrant examinations l
(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 l
and repair of the concrete.- The inspector performed the PT on the welds as required by the repair peckage and the procedure (QI-QP-11.18-1,
" Liquid Penetrant Examinatio.~.'). Scattered weld porosity was identified
{
by the inspection. The porosity was ground out and a repeat PT was-i performed. The final inspection is scheduled to be perfomed by QC inspection personnel. The liner plate areas to be inspected by PT were identified in DCA 20856.
No violations or deviations were identified.
7.
Cadwell Splice Observations and Records a.
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 perfoming 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 Rest,rir s (1bs)
(1bs)
(lbs)
~
0 0
0 0
0 machine on 4/2/85 t
100,000 99,750
+250
+0.25 200,000 199,600
+400
+0.2 300,00 299,450
+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 4
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____.____-_m
l l '
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 +/- 15 accuracy requirement specified by Calibration Procedure 35-1195-!EI-37. Revision 3, dated March 11, 1982. The reference standards were identified as follows:
ID No.
Manufacturer Calibration Due Date RS-75 SLH Electronics January 27, 1987 RS-75.3 BLH Electronics January 27, 1987 b.
Observation of Cadweld Splice Tensile Testing (1) Qualification Tensile Testina On April 2,1985, the NRC inspector observed the following tensile testing of cadweld splices for cadwelder qualification:
E80 08, G8W Q1, GBH Q2, G8V Q1, 8FD Q4, 8FD Q3, 8FH Q4, GAH Q1, GAV Q1, and GBV Q2.,
Each of the above qualification cadweld splices was tensile tested to 400,000 pounds (100,000 psi) and met the requirements stated in the procedure.
(2) Production Tensile Testing The NRC inspector observed the tensile tester calibrations and the following production cadweld splices tensile testing on May 7, 1985: FXD 3P, FYD 4P, FYD 8P, FRD 87P, and FUD 6P.
l Each of the above production cadweld splices was tested to l
400,000 pounds (100,000 psi)and met the requirements stated in the procedure.
(3) Issta11ation of Production Cadweld Solices The NRC inspector observed installation of rebar and cadweld splices at frequent intervals (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 relar installation for the Unit 2 closure was perfonned in the area identified as elevation 805 feet to elevation 875 feet and azimuth 300 degrees to 335 degrees. The installation activities observed included rebar l
spacing, location of cadwelds, observation of selection and remova for testing of cadweld splices for testing, and determination of location of rebars and cadwelds for the as-built drawings.
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L (4) Documentation Reviewed The NRC inspector reviewed the following documentation for the
-l rebar placement and cadwelding for the Unit 2 containment (reactor building) closure area:
Drawings DCAs NCRS 2323-S-0785. Rev.7 22616, Rev. 1 C85-200294-2323-5-0786, Rev.9 22728 C85-200339, Rev.1 2323-$1-500, Rev.5 22737 C85-200355, Rev.1 2323-51-506, Rev.5 22836 2323-52-505, Rev.5 22878 (Sheets 1-7) 2323-S2-508 Rev.2 22772 2323-S2-506, Rev.3 No violations or deviations were identified.
8.
Concrete Batch Plant Inspection, Unit 1 and 2 The NRC inspector used a nationally recognized checklist to inspect the concrete production facilities. This list included the specific characteristics for the following 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 ap ticketing system, and (5)plicable because it had been dismanteled), (4) delivery system.
The current batching is a manual operation since almost all concrete has been placed. The central mixer was dismanteled and removed from site two or three years ago when concrete placement was virtually completed.
Presently, the backup batch plant (which was a backup system for the central mixer) is in operation to complete the remaining concrete placements. This batch plant is in good condition and complied with the subject checklist except for one area.
l 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 nebers and the truck then mixes the batch to be placed.) The b'ades inside the truck are subject to wear and should be checked at a reasonable frequency. The Brown & Root (BAR) representative responsible for checking the blades in accordance with BAR Procedure 36-1195-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 ~
such checks dating back to 1977 when they were initially checked.
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
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1 l
mixing blades was not documented, although the BAR representative stated that the mixing blades were periodically inspected and laboratory testing would have probably indicated if there was a problem with the mixing blades.
Strength and uniformity tests have consistently been within the acceptable c
range indicating that concrete production was acceptable even though j'
mixing blade inspection was not documented.
Otherwise, the condition of the inside of the truck was satisfactory as the drum and charging / discharging were clean. The water gage and drum counter were in good condition.
This failure to follow procedures is a violation of 10 CFR 50, Appendix B, Criterion V.
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 identified.
9.
Calibration Laboratory for Batch Plant Unit 1 and 2 The NRC inspector obtained batch plant scale numhrs from tags which l
indicated that the scales had been calibrated and were within the l
calibration frequency. 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 activated. The records were adequate except as follows:
a.
Scales NTE 766 records do not clearly differentiate between the required accuracy of the scale and the digital readout.
b.
Scales MTE 779 and 780 records show various accuracy ranges for the same scale; i.e., MTE 779 (SN749687) records the following: report dated January 1976 gives 1%; report dated July 1976 gives 11 while the report dated October 1976 gives +/- 0.21.
the above items are unre-The calibration appeared to be proper, however,s actions regarding.the solved pending further review of the applicant correction of these recerds (445/8507-05; 446/8505-03).
c.
Records for scales MTE 779 records contained B&R meno IM-1108 dated July 16,1975, which described a nonconforming condition. This condition affected the water and cement scales causing a 24-48 pound deviation during the calibration test. 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 that corrective action included an evaluation to determine if concrete production was adversely affected.
)
, i This failure to assure that a nonconforming condition was evaluated is a violation of Criterton XV of 10 CFR Part 50, Appendix B.
(445/8507-06; 446/8505-04).
- 10. Concrete Laboratory Testing Units 1. and 2 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 procedure.
The NRC inspector inspected the testing laboratory equipment and found the test area and equipment were in good condition and each piece of equipment was tagged with a calibrat'on sticker which showed it to be within the required calibration frequency. Test personnel were knowledge-able of test requirements and equipment.
The NRC inspector witnessed field tests performed by laboratory personnel as follows:
Date Truck No. Mix No. Ticket No.
Air Content (5) Slump (in.) Temp (*F) 6/3/85 RT-41 925 64013 Req 8.2-10.3 NA 70 max i
Mea 8.7-9.1 NA 57 I
6/3/85 RT-35 128 64014 Req 5.0-7.0 5 max 70 max Mea 6.6 6.25*
57
- Truck was rejected by quality control but was later accepted when second slump reading came into required range.
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 3000B, 3002 and 3004; Sieves MTE 1286, 123it. 1272, 1274, 1136A, 1156, 1094, 1093, 1095 and 1067; and 2", grout sold MTE 1111.1178,1179,1300 and 1180; Aggregate s The following test records for placement number 201-5805-034 were reviewed: (1) concrete placement inspection, (2) concrete placement sumary and, (3) unit weight of fres!) concrete.
No violations or deviations were identified.
- 11. Inspection of Level C and D Storace Unit 1 and 2
~
The NRC inspector inspected all laydown areas where piping, electrical conduit, cable, and structural reinforcing steel were stored. These
' materials were neatly stored outside on cribbing in well drained areas L: rrr___rn _ -
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i which allowed air circulation and avoided trapping water. This met the l
Level "D" storage requirements of ANSI N45.2.2.
The electrical warehouse contained miscellaneous electrical hardware.
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 term < nation l
tool room). Two minor problems were identified and the warehouse personnel initiated action to correct them.
I The first problem noted was that a box of nuclear grade cement was marked
" shelf life out of date" but it had no hold tag. The box was subsequently tagged with Nonconformance Report (NCR) E85-200453 after being identified by the NRC. During discussions with the warehouseman, 1
the NRC determined that engineering told the warehouseman to mark the l
material and lock it up, but did not tell him to apply an NCR or hold tag. TUEC should determine if engineering is aware of nonconforming material controls and provide training if this is other than an isolated j
i instance. Also, the NRC inspector noted a very small leak in the roof above the electrical termination tool room. This leak was in an area that did.not expose hardware to moisture. The roof is currently being repaired.
The millwright warehouse stora e area was inspected; however, only a small number of items or mater als were stored in this area. The overall j
storage conditions in this area met or exceeded Level "C" storage
{
requirements.
No violations or deviations were identified.
- 12. 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 internals by examining the completed installation and inspection records.
j a.
Requirements for Placement of RPY
{
Requirements for placement of the RPY' to ensure proper fit-up of all other major NSSS equipment are in Westinghouse Nuclear Services i
Division (WNSD)
- Procedure for Setting of Major NSSS Components",
i Revision 2, dated February 13, 1979, and " General Reactor Vessel Setting Procedure" Revision 2, dated August 30, 1974. The NRC j
inspector reviewed the following drawings, which were referenced in i
the RPV operation traveler, to verify implementation of WNSD J
recommendations:
I WNSD drawing 1210E59 " Standard - Loop Plant RY Support Hardware
)
o Details and Assembly"
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WNSD drawing 1457F27 " Comanche Peak SES RCS Equipment Supports o
- Reactor Vessel Supports" i
i
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CE drawing 10773-171-004 " General Arrangement Elevation" o
CE drawing 10773-171-005 " General Arrangement Plan" 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 1
criterion such as centering tolerances, levelness tolerances and clearance between support brackets and support shoes. 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 procedures.
4 g This is a violation of 10 CFR 50. Appendix B, Criterion III
/
(446/8505-05).
dr~ DAME-2 b.
Document Review
~ -- --
The NRC inspector reviewed B&R Construction and Operation Traveler No. ME79-248-5500 which described the field instructions for i
installation of the Unit 2 RPV. Requirements recommended by WNSD procedures were implemented in the traveler. Worksheets attached to ""/
the traveler showed the RPV to be centered and leveled within the 5
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 38 of the traveler i
indicated an as-built clearance of 0.012 to 0.026 inch which exceeds j
both the original and revised tolerances. This conditipn was accented gI on the traveler based on Westinghouse concurrence, and'There were neither nonconformance reports nor documented engineering evaluations to determine if the condition wa's acceptable. This failure to document nonconforming conditions and engineering (deviations is a violation of 10 CFR 50, Appendix B, Criterion XV 446/8505-06)
The NRC inspector reviewed the following receiving records for RPV hardware and found them to be in order:
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 and 26 0-Rings o
Deviation notices and corrective action statements
_ _ _. _. - _ _. _ _ _ _ _ _. _ _. - _ ~
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r TheNRCini tor reviewed the following completed travel rs for internals it allation and found them to be satisfactory:
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ME-84-4641-5500. " Assemble Upper Internals" J'
o ME-84-4503-4000, " Install and Adjust Roto Locks" i
i
,i ME-81-2145-5500 " Ret [rque UI Column Extension" 3
o o
RI-80-385-5500, " Transport and Install Lower Internals" ME-84-46'17-5500 " Repair Lower Internals" o
o ME-84-4640-5500, " Assemble' Lower Internals" c.
yJsual-Insystion e
I At this time, visual inspection of the internals by the NRC inspector was not possible, and inswetion was limited on the vessel placement to a i"
walk-around beneath tie vessel to inspect, the azimuth matEfrwi and for construction debris between the vessel and cavity. No problems were identified'fn this area.
d.
Records of QA Audjtsgr Surveillance The NRC inspector requested TUGC0 QA audits or surveillance' performed by TUGC0 of the Unit 2 RpV installation.- TUGC0 did et make,available any documentation of an auditine surveillance which/ evaluated speci-fied placement criteria, placrxent procedures, har.iware placement, or as-built records. This ites is unresolved pend'mg a more comprehen-
,~
sive review of these activities (446/8505-07).
f No deviations were identif tad; however, two violations were Mentified and are described in the above paragraphs.
- 13. Reactor Vessel Disorientation On February 20, lyly, the applicant reported to the NRC Resident t
f 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 degrees.
Initially, Unit 2 was to be a mirror image of Unit 1, however, a design"-
change was initiated to pemit identical components for both units'. The-design change was implemented for the reactor vessel,-but not for the pedestal support locations. The problem was not considered by the applicant to be reportable under provisions of 10 CFR part 50.55(e) since r the error could not have gone undetected.
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The deficiency was reported to the.NRC Office of Inspection and Enforce-ment on February 22, 1979 and during a March 27, 1979 meeting in Bethesda, Maryland, the applicant presented the proposed redesign and rework proce-dures for relocating the pedestal supports. No unresolved safety concerns with the repair were identified at the meeting.
During this inspection the NRC inspector reviewed various documentation relative to the misorfentation problem, including design changes and the construction traveler which implemented the repair.
The following documents were reviewed:
o NRC Inspection Reports 50-446/79-03; 50-446/79-07; 50-446/79-13 o
TUSI Conference Memo, dated March 1, 1979, H. C. Schmidt to S. Burwell (WRC 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 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 RPY 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 identified.
}
- 14. Reactor Coolant Pressure Soundary (RCPB) Systems C'
The inspection was perfomed to verify: the applicants system for preparing, reviewing, and maintaining records for the RCPB pbing and components; that se' ected records reflected compliance with NIC requirements and SAR commitments for manufacture, test and installation I
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,7 of items; and as-built hardword wa' adchuately marked and trueable to s
records. The following itev were rando:altselected and inspected:
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PressurfA}er defetg Vriva~- Tah'f tim was ins ~pected to the commitidnt a.
stated )TF5fCa5Te T'@l ifnfch includes ASME Section III,1971' Editfog through. Winter 1572 Adden'de/ Yalve S/W N56964-00-007, 6hich k,4 is installed in the B position was bnspectini. The following records wergevviend:
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1WReceiving Inspection ReNet No. 21011 3t 9i
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o Code' Data Report Form NV-1
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alve Body
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The.'v 1ve was in place, however, installation had not been completed; therefore, the hardware installation insp+ition densisted of verifying that the item was traceable to the records.
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b._
CYCS Spool Hece 30!!9 Requirements for this item are., stated in s
ASME,Section III, lW4 Edition throuSh Susse? '1974 Addenda, whhh is the committment from the FSAR, Table 5.2-1.
The item was field 4
fabricated from bulk piping ani purchased elbows and installed in the j
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CVCS with field welds number,1 and 6 (ref. BRP-CS-2-RB-076). The i
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i following records were revfaw4dn BAR Code Data Report 1
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oq Field Weld Data Card g
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oa NDE Reports t'
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b QA Receiving Reports for piping and elbows 1
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>c Certified Material Terst Rycrts (CNTR)
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s The installed spool piece was inspected for weld" quality and to i o verify that marking art traceability requirements had been act. The s
1 item had been marked witti the spool piece number.(3Q1) and the BAR 1
drawing number which prbvided traceability to the material 4
" certifications.
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- Loo? 3 RC Cold Lee - R W resents for this item are stated in ASME,
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' sec ;1on III,1974 Edition 41 rough Sumner 1974 Addenda, which is the commitment from the FSAR! Tdble 5.2-1.
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i consists of a 27.5 inch cast pipe with a 22 degree elbow on the reactor end, a 10 inch 45 degree nozzle, 6 3 inch nozzle, and three 21/2 inch thermowell installation bosses. The following records l
were reviewed for the subassembly:
l o
QA Receiving Inspection Report No.12389 Westinghouse Quality Release"(QRN 47523) o e
o Code Data Report Form NPP-1 i
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271/2 inch line CNTR o
3 inch Hozzle CMTR I
o Field Weld Data Cards p
o NDE Reports j
1 (1) Sandusky Foundary and Machine Company test report for the cold j
leg pipe certifies that material meets requirements of ASME j
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Section II, 1974 editions through winter 1975. Southwest
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Fabrication and Welding Company code data report NPP-1 Form i
certified that the cold leg subassembly met requirements of ASME Section III, 1974 edition through winter 1975. The FSAR coauitment is ASME Section III,1974 edition through sunner 1974. This discrepancy is unresolved pending the applicant's evaluation to determine if material nonconformances exist l
(446/8505-08).
1 4
(2) 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 l
Installation Verification, and N-5 Certification." Procedure CP-QAP-12.1 requires that a data package to be used in the test, i
be prepared with the test boundary and the additional following data shown:
o Base metal defects in which filler material has been 1
l added, and the depth of the base metal defect exceeds 3/8 inch or 105 of the actual thickness, whichever is less, o
Untested vendor performed piping cin:umferential welds.
o Approximate location and material identification and i
l description for permanent pressure boundary attachment I
with applicable support number referenced.
o; Weld history, which shall reflect weld removal and/or weld repair.
l
..]
,:r 4
...,_,,,,...,c,9
l i
l l l
l l
i 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 boundary. A handwritten statement on the drawing" indicated: "No major base metal repairs could be located and "No hangers with weld attachments could be l ocated." Welds performed by the pipe subassembly vendor, including the 22 degree circumferential weld and the penetration fittings had not been identified. The following items are unresolved pending further review to determine:
o If "no major base metal repairs" was based on a visual inspection or on a review of vendor and site inspection and repair records.
o If the shop circumferential weld attaching the 22 degree elbow to the pipe assembly was inspected during the test.
If welds for penetrations into pipe assembly were inspected o
as procedure CP-QAP-12.1 does not require identification of such welds and they were not identified on the drawing.
The above issues will remain unresolved pending further l
eval:::.t?on by the applicant (445/8507-07; 446/8505-0g).
l l
d.
Personnel Qualifications - Personnel who had performed selected tasks i
were identified during inspection of installation records. Training j
and experience records for the personnel were reviewed to verify that employee qualifications and maintenance of records were current and met requirements. Names or codes for five welders and two NDE l
examiners, who had performed tasks during installation of the items being inspected, were identified and their qualification records reviewed. There were no questions in this area of the inspection.
No violations or deviations were identified.
- 15. Special Plant Tours (Unit I and Unit 2)
On May 23, 1985, the NRC inspector conducted a tour of selected areas of Unit I and Unit 2.
The group consisted of one NRC inspector, two NRC Technical Review Team (TRT) representatives, two allegers, and several TUEC representatives. 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 analyzing this information and will decide what action, if any, should be taken.
/
l Qo h h
c/M-4
s 1 During this tour the NRC inspector independently identified a 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 stated that areas on the wall were reserved for pipe hangers GHH-51-1-SB-038-006 and R1(?)1-087-X11. These stickers were dated 1980. It was not evident whether hangers were missing or none were needed in these locations and the reserv.a tags were not removed. TUEC representatives were unable to answer the question immediately. This item is unresolved pending further review during a subsequent inspection. (445/8507-08).
No violations or deviations were identified.
- 16. 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 building.
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 activities.
l No violations or deviations wue identified.
i i
- 17. 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 vere complete and up to date. A total of 183 reports have been subiitted to date. This inspection period one Part 21 report on Diesel Generator 011 Plugs and two 10 CFR 50.55(e) reports on the Equipment Hatch Cever and SA106 Piping (light wall) were submitted.
No violations or deviations were identified.
- 18. Review of Violation and Unresolved Item Status i
The NRC inspector reviewed all violations and unresolved items reported l
to date to assure that NRC and TUEC status logs were complete and up to date. Two hundred nineteen items were reviewed. In addition, a trend 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 B.
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; X.
Inspection; XI. Test control; XIII. Handling Storage and Shipping; XVII.
QA Records; and XVIII Audits. The most significant trends were Criterion III, V, VII, IX, X, and XVIII. Also, a number of violations occurred with respect to 10 CFR 50.55(e) items.
(
~ + - ~ ~ -..
m.,.n.,
/
,,\\
21 f
These findings mainly pertained to Unit 1 and related closely to trends identified by the NRC Technical Review Teain TRT. These trends will be considered during followup on TRT findings. Also, Unit 2 inspection emphasis will consider these trends during future inspections.
No violations or deviations were identified.
J
- 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 3
inspection were discussed. The applicant acknowledged the findings.
)
I l
l 1
3 l
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i 1
1
_m.
_.o
Mib t N i
WESTINGHOUSE PROPRIETAF Y C 2
REV. 2
)
- '* 5N D'"Se
)
Westinghouse Power Systerns Electric Corporation Company
- n, l
e.rs:.prerseama n l
February 13, 1979
)
WESTINGHOUSE NUCLEAR SERVICE DIVISION PROCEDURE FOR SETTING OF MAJOR NSSS COMPONENTS 1
NM REF:
NSD General Reactor Ve'ss'el'Settin'g Procedure
?
- ep w
.c hc.
pas. ( T
- d 3 x
.b r
u
..- ~ M rb. e,.v W ! ? h w Accur.sev n cettino ma4or Nuclear Steam Supply System RSSS components L e rescensibi' ity of the custnmer's erection enntractor.
It
~
i
,is the so
. is suggested that said contractor provide a "tf Ted settin n. h y
.for review by Westinghouse NSD site personner. ;, ne use of this document 3e and/or our specialists, as consultants, in no 'way shifts said 14 4414+y l
to Westinchousem Good erection' practices'must, prevail.
. In both orientation and elevation, the reactor. vessel (RV) is the key l
RV elev-
'g
. ;. 3 3,y, component' to.the setting of' all,other ' major NSSS equipment.
J ation is detennined for optimum fit-up'"of the reactor coolant (RC)
' " ' 'a ? '
. piping to.the other components.
- . Wd......? b
.n',The folloWing 'is 'a sumnary 'je V 4...
,.p;W#a'n~.d suon-+.d ' seque a :.i n r.4.,%..,+
1...
~
ofjasic requirennents.
' v forQnsta11TnfEe'Ef, fuel transfeF system'(FTS) "RC pumps. steam gen-pres,sur,12er,, pan,d,,RC piging.jg g%g'g ep g
dhera, tori;(S{j
,g r.v.
e
.... 9 %.g. QUI,S.I.TE.,g t-Azimuth and.e.,a r -.elevati5I ap.,,,..
g.
,a.:
enc g...;; g' PRERE M thin'-the' containment.'MI..
o.>. v m. sof. loci,o c-3 'r;.9.:must'bi a.VailaM,i dh
...a gn a. pup em.pu..
3,,9. u.g.
c i
' J mc s a w.
- u v
',.,,.,.y 5,.. Q'{L'ay out.the~ q h'eoreti c. a u..=.< iriment t
al conta
' axes from the abwe indicatM benc.h marks'."C *v.. r; q. 9 ak p ; % lb 0
g.,
C,:.q
' MThe reference pot.nts used for, setting.the ' vessel. suppo
- c.
shall be the cente+11ne of the suon' ort'and the* horizontal 97.4
- T M -
- a..,as,
.;c, y ; n o.tJ.n,.I ve.. ;.
' ' bearino olate of tie ' support.d r
w.,
....4
~
3.
The setting tolerances'for 'thelsupports shal1 be:
- A l
- h:'.!' h f.
.%.p., ';p j.
<: g,
- f..
,+
n
.' (
'C The supports shall be grouted in'.
The reactor vessel support cooling channels shall te hydro tested.
$OOBA00y9+~
IV-1-1
==
\\ REY. 2 WESTINGr10USE PROPRIETARY CLASS 2 I
(
./
/
]
~
Westinghouse Power Systems w w se..:e o.. m Electric Corporation Company n,2.n
- s Pgi m g1Fe etyt.3aa U2M
- 2-4.
Within the equipment support system's design flexibility, calculate the opH mum RV elevation from the RV, RC pump casing, SG and RC piping "as builts". This optimum elevation is.the mean horizontal plane required to average any variances
,in the equipment nozzle elevations for maximum RC piping fit-up.
During this exercise, also determine,the.SG and,,RC pump, casing azimuth' location'and orien'tation for the'same re.ason.
_s, 5.
Insta11' and functionally check out the neutron detector L ~3 E
f *,.c.? ;u J t N,s t.qcw A e52 % rib.Pde4 -
positioning devices. L 2.-
6.4'Due'to" plantla'yoilt,'it" day *q.gr.f1PE'be' reg'uired,or'at'least'b i mJM :7'. '
i' 3 to position'iome E
" the shield ' all /all 'th:e'RC ' piping adjacent to"the~~RY throt;gh ~
w sleeves" prior to" vessel settiri
- Positioning
'the bottoin %unted instiume~ntation c
o'f, Med' ribc~e's's'ar'.T**4 '"$.pM6 P'" ' pipe 'm
- j.? '
dee f
O8 4.
/46?g?
i :.. $ W.G.be,' checked, ~b Y
'N d.h'A ! W T $
2 "'
i.
nda.
The reactor.' vessel card'(
.s ma.in s must A
i veritied,"and newly nia'rked if. Markiri "1 Ifo'und 'in'c~o'rrect W ' W.p'/
l.
'or 'nWt Nisibi d'e 'M 'M RA.%M"
? M ' W'..
W. %*$.w
$[ bf,i'. a..ptef Q* Q !.. J i 4 W ; &.. % &
.. h'The.reacto ves,s e,1 -
,;.oca,t,e,d;ciycumferentiall
'.,','"*i*
- E*
h4k t
v$
Set the$4 on
<g r,, e
..m
.e-
'~
.q RV,t re etermined. pt cium e eva ~
W.pe - l' :
7.
f.V' eference la6e ele atio
- 1evelness['
.,.h Ok'.
W Wfb* dst
'd fj 's't" he'ma'tiri ange
~hufTace.v. nem v'able Wot'ecti'i onr plates allow' access 'to %:,hi.
the. support'1 edge,for..determ
. actual., ele ation~ and '. levelness'p-
~' du~r'i sett n'g'd 'Liveines'so gi.i' bei'withinYf44 M fN i+
O... W f'., "
2bMP@%.hM b U
,.kF
,.. ys
- 1. "
- c
/
%'5
' parallel to the,el.guel},*ti1)ent the.fWM9F eidt W*
. t.1 same containment theoretical < axis,4s aligned ig.
.. proper.ma or.,7ax s c
,vess tg utilized for layout of the FTS 't'ube'and ' manip'ulator dNne guide rail.
'i -
. Scribe or punch marks are located on the, main-flange vertical f
. surface at the 0*, 90*,;1_80* and 270* axes, accurate to 3 0
...,,J.
- These marks can.be employed'forlorientin the vessel.i p
r, d.he tR refe'rencing y the'four' barre,,i s;best ocated,circumferen_t,jally;hekeyways,,a "A;J.
pays.{
6* covers will!also provide'
.$4% :.
V Most. future!ves,s,e
'tothekeywayspf}{protectoQ,Qy.{.y.,;.jf.f.'f "'(:":W Fy
- r. -
8 8
,y
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- c.h
~
WESTINGHOUSE PROPRIETARY CLASS 2 REV. 2
. " "'5r' 'O'
- Westinghouse Power Systems e me.'..
Elecaic Corporation Company
- 4.. ;fgs73 eve.r4 pr; i.
)
1'
. d
. k. r,.;,.
'.va.,,. : :.%.. >. s, r, l
g 8.
Final set the vessel per the referenced " General. Reactor Vessel Setting Procedure".?.'The.Lvessel should.be3 nsulated during i
,the period of.'. time.'wher'i"the' suppott, bottom,'3' Mijs}ag;,b 1
yp 4Qf,f g. d,7
- 7...
-w W.- (c:.+ MM cpy(ar%., (..=,m-e-rw-machined.' ~ -g. q ;<.9 chi.tect engineerssupplie% W J GP d Hilicoils".
n.'
9. Rough set the SG's.
~
installed within' supports)~ and RC p' ump casin'gs','in cold P
M position, to..the..elevati.on.
i th and orientation *as calculated,
i
- P
'during determination of ;the,az muRV 'opti,m,'uiri e)iva't'i on".fo'r C.
- 3h pipe fit.up..p Axes. marke.d on' SG's7 a
, shop use,
l
- m p and thus,;should only.sbe..enployed 'a,s.'.a uideMhe'radia,1 dis-l "i & ctance ;from ithe ;RV.cente aid ec u n'
nozzle's.should allow ;
$.M.e 4/32 ji.
I J:!O t.(NOTE ftper:stepTI a
.? ittf und % (js G U Y a
k.f I
cogdancewit n; gt 4.,c.itJGeneral Informat on gp,g.jpj
.)
% ' trMspt65ni6 Tit 4)tti:iWI ps ** ' Wj
' ".'a.}' pre-deteim, genera'torsjshall be set,,.
rY" points shall"be ft S,The. steam 4
ined Work p61rits 23The'se
' 'onent g,1 ensions' and. > -
%the.: points shalli ie 7{5s iiiJt"c r.detenninedfusingith
.thd mum' react'or vessel H.
oMW 7
'A 9 Mri e
..w. A, 3 5..,'
2p i "A.e;4 W jje.levaty n.o m n u m scw ma}c Lg.o.
,g.g.
.n S:
t for,the elastic
..:. Consideration.mus e g v9n-cco DtieM. cdna,eflection 'o6the SupportM] s '
g rf.y;d6'%
/dc lElevation".should be m.
Q t t W.'2 h p 'dj u[2ted40rhi,spfpc'ti W
Ani 9WVA l",Wr.te' Mhc.i P.fior,$6mree.. w:M Jg sonic
.n J
to set ngthe.suppo.sy e'
qu red. ultra-&
ould'h' ave
- been^perforced '
.j
. iQ.)pi.., y A
4
'd '
.'W..n,! h ;p+:,p.and,a[cesp, e;d.,..g;;-.q.y.T,g,3g t
3 f
gip,e -y _
- ~. 3.,
.9
~d.
To insure that~the ua ted on the :,. /
t!. 2,..Lsuppor,t system..some load.measu'ina deii,ce,'such,as strain r
m) ;,.c i
., ( r. :. t. s :.',<., gages or. some other; method 'sh'ould 6e'aoreed on!by ' cons truc-o.'k.? f.1 t, tgon. ',.([f,,
, '{. [ '
'M.
y
.).
w~
'J
. w.gw.(g 7;..gy:s u
,.., 6..<..e. g $ (Q P..i:f:
M.cffgyf
/~ %
4 ' '.:
i M.1..
u n..- w
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a.
c
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~
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~m=--
,,-.g
p REV.
2 WESTINGHOUSE PROPRIETARY CLASS 2 The pressurizer may be final set at this time and is to'be
]
10.
vertical within one-half degree.
It is advisable-that the
./
plumbness'be less than one-half degree so as to' facilitate.'
.L,
.r fit-up of auxiliary lines.
7 Utilizing the RV nozzles as a reference, check'the 59 ano 11.
RC pump locations via instrument and adjust the equipment accordingly. The SG's must be vertical within one-half degree whereas the casing's main flange must be level across the 00 to within. 0.125".
- It is also advisable that the -.
olumbness of'the SG be less than one-half degree so as to L,
facilitate fit-up of main steam and readwater piping. e....'?,.,
(
NOTE
' Presented below [is just one of several.......,..o.s.n:.
.?*:'.6.. t.x. reacto
.g
~
i f.^ pjpe' weldout' seq' ences successfully employed.,to date.1 u
. y.. g.; x.q. 9. u.>..pg,qp y.use ggg g. ju n tjr,
... 3
- 12. Aligihot'and,~ cold leg pipe','or% pip'e assemblies"(with, stosi ; f f#W
. [.f.
i
~
between'.dnits and ~es'ta 11shMreference marks tto detect p0.~
I valves) k, ps'ing' prosier ' welding" technique.and'sequ'encing t (@C
.?
pipe'dra
,to maintiin )1p'in in aligned ' position. attach. pipes to the'
'.e.i.,
~
t i
i lignment.
x 4. Elbows' can 'a s'o 'be ' rot'ated 'slightly1to ass st e n agitp
". RV If,. loop'stopTv'alves*ard1to tiCins'tal1edppric,eed :with w
. eggj, n-n.
f e
J
.! W,'
of' valves.to" pipe secftons'alriady'attachedito the'.vesse1 N,
p6
' thro [ alignment
'and then ' remaining pipe pieceit6~.vsNe,'. fig tt ugh p." M' V,;
ine c, 7 i
be mainta.
..( reference'udr.,injthe. assembly.austn. e, f u.a,,
,ofeach; item J.
el
. 7 Q.u..c,g&.
P 6
ks aslindicated l
c.
agnede..
- .< y.
le is]16ng4sicen, joints,.. a-
..,... +
. p..s... w
' ~~i"Many"odeiloop.)s] din...N u. w.,.:.14..
l severa s
y in M.T
. rw y, e,,, g HOTE - Simu,ltaneo w
Ml d.%'
f terlind) acc'ept
'A" W
. N4alignmenttis maintained -throigh"estabilshed 'r'sference'
..f tf?so' 1/32" clearaVice must',be'.
~
..y,g~ h' alloWe.doneran addhionaly4 *E m
4'#
d or* at 'the valve g m %p,c 4
. s e r: 6.-
A., e. q
..rease
/
w g w,o,s,.i,,,,,,.
. ;,Ql;g 13.T.g.ev..w. Weld sh.inkage.
ap, - twe
~5G s'and pu.en..ymp g 11,.acc un,
n
,m, oving"them;l 3..,,g gyt;.a.
. 7.,the.,insta led. pip $he_,,equ',1poen)pmb,1ies a.
e.,or.pjpe asp
.n t/.
~?-
- Adjust, Plocatio'ns dl3casiogs' yM 1fsch%hitJthe'b,ri.ginal,1/32a') line,.throug
$the connectfng no:: ale 'centerT.Wairf RV r..
4an inwar. direction.' parallel N'a'nYc51"d7egsToIs a'nb $defcIsing f
bh "NM.E"E'khEb"*Nf'fks h hh Nf O
M
- 14..ded.
' 'not to. draw'ahy major NS$$ eqdipment <otherj'than rad' ally ::
.'inward.
weld.shrtrik'i". rom their 12 J W "..
$'.,. previok,towards ;the RV due.to'rWlf ' "
s W
g;...:..,
a sly set' positions?
' c'. @
% ny pjuT;c.G 15.0Aliin&DM{.
ment.and mary c n ead oupet fa'ellsitoiSG'j'y.]weldoit..of the" ime]~duri can:beidoneanyj eciding Wweldin'g% era'tlois @Be' sure Katthfir ree "tield pre forms a horizon.tal plane'.@p'" yf g 9
..g~.yW.;;,g.
,ye * /?..W
)
yy
.n g
+
ilZZI_~EZZ J_M'i r-~;_ p_ep..
--.~
'i f 7
REV. 2
" 'STINGHOUSE PROPRIETARY Cl.Ar p l
i 16.
Measure horizontal and vertical RC crossover pipe clon re dimensions to within 1/16" and sulait through the Westinghouse Nucicar Construc. tion Department Site Manager.
Said dimensions il
~
are not to contoin on allowance for wcld shrinlage. The.
crector is urged to separately advise desired excess material for shrinkage, otherwise the closures "as received" will contain an additional length per Westinghouse standard.
A three-week advance notice of data submittal date will considerably decrease closure' delivery time. :With said advance notice, the first closu're can normally be shipped within nine weeks of data receipt. - An additional. closure N:s:-c.t,
..)
can usually be shipped during each of the fo11osing'wcekt'l. $ 4 if data was likewise submitted on a timely basis'tojllow 'ly5k..)7.
i fourweeks,shoptimeeach.ry.h..y 4: 4.,.-l$.; :f Q t'y.; ; "..,'.
1 NOTE - Closures are normally supplied in c.:.y b,.wj,~;<').
s~,
.as
' wg. ;,,.
two ~ sections.per,. ll T '"
y M.. ? e/'
loop but if. plant ilayout permits, the customer.may ;
initiate an agreement with Westinghouse,WRD-PWRVphP;.;f:
M.-
W '- @
i O
SystemsDivision. Projects'foronepiece' closures.%r>l'E
!?.
,u-x.t.
.% a - ::*4.Qs %6)M.'W. i0%%d 17.
Utilizing weld. techniques as employed for hot and cold,1egf.
- y,.,
l
. n.y i
.. piping fit-upf weld out the surge line~ between hotcleg nozzle ~and' pressurizer.Q.T'
!by.r@.#f-i.m i
p%
. l W.. E R tt W @, f f, a 6.
Rig' RC pipe'Q+s:u:rle:s..w't.yi
.e : :.U same fit-up techniques, weld closu,r,es %o.SG ells.
)
clo
~into position an, again' utilizing'the
- 18.
ings. h,<.o.t,4
.l ':;MWtwo sections,lMWriN%.t6,hW's' teel. weld
- 1. $.
%-i~.cf@
~
and.if.~in W'.Y[?m
~"
L.?."..'
a M W'IE'e
)
Install equipment support seismic' restraint
- 19..
w G. V S t.'. g p. g. u. M u m \\..:mp;w
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si. 4=.l:
A.
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.y e:. wr.e.;;m.u...a.c v
- s. s wct.
w.n
.. w.u
- f g.fh,...
.Wd M.c.4 y g.;4. Wg.:d $M.' 7. <.s1.g.
h, e.l. l0&..h,a$.d;..A.53 ; n mw$ $u h
.k._f.f..
.'ih:k<kh.,s?W::~.:
k
.3 c..e. Y.* ? a
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- g. ; g.b;?.$
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W
.' K,. 1
..,$,.k,&3....
+ NOTE &.f.,..'Tl e.y v
V.=
2
..q. ;'
v.,.; p,,.y *p y.3...,,.,.. w..
.3
..v r
.m f.
-.v.,.,,;
Only two' welders are required to' start welding the fece, closure leg center weld. c' Welds at the pump casing 'and stesm~generatcMell:ahm "%
should not be started until sufficient centerNeld shrink'has'ciccurred e J'.
to accomplish proper fit to the, pump and s, team. ge.nerator.,
ydi.
"Q'l (,
3 9hhn.N.g;$$
-S, $..,., '.
..,N'.\\g.h.,i,f,h,bh ~
.~...hb
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n 4.Tp;....E ;.J h'..'
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- 4 IV-l-S
__/
- WESTINGHOUSE PROPRIETARY CL.
r2 Rev. 2 8/30/74 1
J MiES
, _j.,
GENERA'. REACTOR VESSEL SETTING PROCEDURE ',
)
- 1. ;.e n.
f
~
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.. ;r.
.. w-
....n.
~
.g.
g ac. g.,,. ;,, g g
,g IJ Objective r ;,
. :... u.:... r
- r. *
. ~ wu '
- To provide a safe and efficient method for.dnstall,at{on.. orientation.
< and leveling pf, the reactor vessel.v.The reactor. vessel,must be leveled s)
..using.the. intern.als seating fia e.as"the.Batum?planefor))eNelingto
[Thevesselazimutht61 within a tolerance of 9 0_p'mus~"~
gg orientation and eleva'tTon e ma a-ne to -
.of,therea.ctof.,coolantpipingandthe, fuel,. transfer.4,quig'qt.
,y 4
.,: v i i.e : n y:. f ' z:;<;'.% ~ : ).a sp... i:..*: s "G.:Q. ni, %f 3 II. Conditions
, e...j n.;. g.sdi,.
i Thereactor.vesselsupportshoes, shims..leveljngscrews,andmis-
- A.t
, m.cellaneous. hardware are on hand..,..,,..,.,.7 psy,g..,j g h A calibrated jig transit or t'iltin:g level with scales o?
l
<.r - p :.m n c.
.s :. s;D : n t ~4.dt
- u... :..
vu z
- s..
r targets B.
The level.has a more sensitive bubble i
d...:is available for* service.9 transit', but close' focusinicah bE;abiilem.^f.
~
3.J Ethan. the jig'ihould,be, chekked,,a't,the,(.,sg.'just;prjoy,to
-)
c?
calibration,,
,5s.s. W y.m$m 4,. 23g;,.
setting the vessel.
A ;..
gg
'I ' '
1II. Precautions n,f g.q; h u :i.* N.1
.' nc ar..M :M rid tritt.W 9 '.r4 J Wic-d W w ln: aN.- a.' dropping of Extreme care nfust be taken o prevent. con pat on a A.
tools, etc., onto the vessel to vessel'hiealf ma'tinji'Tufface or.into s covers ve bee '
ved.(.
... Q the, vessel..after<the scale
.g e hik --l j
i
.,. y By. ' echeck'.the. vessel :.lff$1'ng fixture to v 3 nn ' *' er
- s. m 3 4. p.
N-..
.JNM tin r
R
.l.'
- f tightness prior;to. lifting Mf Mesiel." @ 's T
)
., pads are t ht..jnga,,in,st }the.
f,f,la 3.When vessel. shipping. skid is;used;for..the~upe ng Dperation upender
.,. -C.2
. J 'sectiori of 'shippingiskid per. tite' reactor.9este13h.ipping skid drawing,
' verify that the vessel to skid retaining',)trapf[& bottom
~
assembly is tight.
T.WQ,
-$iy,%._.g.
r g
g<.....1;,
]
IV.* Instructions _ p p,
.,9 pg.,
m_. n, h., ;
.,.r_.
n.
.. s.
l t
upport pads..
A..,.,... Set.
. g. m o,a,.the.reacto.r.; vesse,suppor..,.s..
o.es.
.s. rc a;g,,g p.
p...,
.. p
...p,;t%.... r.<
....c...
I ns ta114 the J1evel i ng screws,j n.the,m, w, threa support pads.'.
....B.}" The it'oW of 'the leveling'screu,'"which'de iott r
s al oximately
.,,,spannerwrenches,d,hou'IE
~
"ietipDFrtacis t
. :.w.o a..
,i.
v.0t.. I 1' D v-2-1 i
'~--v.-
-v.,
3.
,,g,,,,,
WESTIN" HOUSE PROPRIETARY CLASS 2 i
Rev. 2 8/30/74 WNES-SD NPS
..) I GENERAL REACTOR VESSEL SETTING PROCEDURE.
C.
Set up.the,iig transit or tilting level in'a convenient area close to the vessel (but not on the vessel cover). This must be such N
that accurate focusing is possible at all scale access -locations.
,/
l D.
Set leveling screws with tilting level and optical: scales at all, f
supports,to.the established vertical target elevation. The leveling screw'elevationWto be' established by' the field 'from J
actual as-erected" iGpport"sh'e elevations'ahd reactor vessel o
' ce.nte.h.kn.e,0 U"ItibnN'.1,"th g.ch.:iev,id.1't.n(Qb.:a
."."as-built" diinens4ons)such'th&t'.the'horrect' reactor' e nknwo ;....HtR '..:
1 f1 4e'a 9 @'
s v.
v.
. a.,
u 0.
'E.
Risiidi' sci 1IdibisWoveEs fN#th'erMa'c'to'r'Ns's'eFshipVing' coM.-
i s
Level readings are taken on the. internals seating flange through
. '."<. g. fy'.i.- g Lph.E2!!Ltj$j,p :"l,..
the acces's ports..
.c..n. : r,, :...q c %.,.,.
.v
- ' Notii fit ls$icbinr'nendedMata' tiltingleve1&iusedto i; ale' level' i
. I
. readings due to the rather strin' gent'levelnes's'teq'u'irements. m.i.
l m.c%.piritlevelsa.908Sh;H1re to be attached to scales for verticality control.
1 S
(
+02tusu. git.yegogn.A.,.8',..,
..m =%;p y N
1 F. $?(owePtha'.Feact6rf viis'seltf the"p'oin't' of 'c6nYa'ctMth the # eveling C idt' 9scress',ilinf sc)iri 'thatihevnsel',We,1gh't Hi' f allT,fullysjmposed onk.
t.ik'ini lif - (.f,f.f.{Jyl*Q.* *
,, #Sig(."p,%ei[
~
cr'ews ar
. the'1ev rews' befo're' fit'jisia's'sured 'th
.." C O contact per. Step G.-
,. f. : 2 g
4q
.c Iti/m.y ;,khec nj fthe' ' ".l:. "
- .w. ;,,
N6tei, UsingTtEansitt' keep"the$issel orientat reactor vessel axis marks with phe cardinal axisllocations"
.the. M
- p+
Ms.(,fpp;e.Ni/cViree.*'W OM.S. avec >aa2.emmems.?
~. ao 3,.. j D k.g.wo..Aeaa w ct isrtsvle
.i (i
' G,l 9ta fse*aYY'e'v'el'i ns'criws' Na't$r'eN.nMJ r o
- i. swig Nu [b "
unti1 'firsi'hontact toined.f Y. @idegippthe'vesie iupp*o'rt" pa'dsjfR.3W:
W%?Q g Q p f. M :.,.
4J.,&."i.:y.:1fR;*C.
5". per foot
d
,e.la,nge, reactor vessel to withirp,c$olerancei
. H.
' Level.the
~
s
' Nf f diameter pf YaTsfng"the rea tor" esse 1 4n
- dfusting W,
J '
thf 6 d cal Jeve11 r
84 '@9899Wrv'. to th level'ing krewstfRei h'e'ckTa'11tirstion Ya~k'e 'h'p,.that no 'er ors %av;e bien' de'$f tffvesiel s verify fp,heQ _
~ y.g...p~n.v.y a M3.g.
... :. r.
h et'glif.'inessureine$t's betwkN?the.'uride uppoit l
"I. h:
bearing surface?a'tVp#'pioints.on each
. support uhd' record.". W;h
..J,.) ads'and..the, support s oe
.y.W*y.*?M.*%.Q}ssu44ng;;}j%npnQ.y,,p
(.
3. ;, '
y i
3,..
.NoteY Telesc6;iing g' ages anil'micromete(sef'be~' ijd'f6rttadng.
3
. :.tnese' measurements ~ ? --"
W
~
r
~ '.
N.
9MWst/>CQMM' pent bolt
./
, '~;. J.q8.Machinel I raa hi ini at s'
~
i W
d
'.7.b bottom ~. 4 *f step'.I P.. p <...,*...'d. are bei 1d.r.4.p
?
ir ' W.
P N,.<'t,
.v Note: cit
,finishedmacg th gnesse p.
l a......tes s
- $J l
not be'1ess than one (1)' inch. j' '."4..y n.
O' S$.'p :..;g?Hf.$.,.g:.
7,,,j.(j.ym,
- y.pik ^_r*g( ^,.-
mt 4,
c
- l; g.. $ ~2 2'
- I
.m.
..mw.m.m.-
9.8,g
%g,,,,,,,
k2 WESTINGHOUSE PROPRIETARY CLAMS 2 8/30/74 i
WNES- 0 NPS 5
}
GENERAL REACTOR VESSEL SETTING PROCEDURE (continued) 1 K.
Lift the reactor vessel and back off the leveling screws until they are.375" t.010" above the be,aring surfaces of the support shoes.
L.
Insta11'the fiolddown nuts.
1 M.
Insert the permanent bottom' shim plates. Do not allow the bottom
)
shim plates to be held off the bearing surface of the support shoes
~. ;..gth by the leveling screws..,,. %.,,w.
6 4.t N.
Coat t
' top surface of i
.. y.f 'tls.: r. ~W,c. s :...
1.;.
l
-f osition on the shim plates.
. Lower the reactor vessel to a ceritered 'p&w.
0.
i 6 n....y W
~; Q @.,
j
..e P.$. Lift the reactor vessel and check lforfsErface' contact between;the
' top of.the bottom shim plates and the'.'; underside of the' vessel. bra
- ..~.p.-.;yes..y;r...
.,;. g...'.v:4
. g ;,,_.p;, ired between'the bottom
..e
. Note: A surface contact.
is'requ j
~
shim plates and the vesse rac e s.
-machine'or hand dress, as necessary, to obtain required contact.ar.eayd.gyf),s.D,:&;.a
' @ um dh,
' % f47.tyVr.gtg4 %.d N--) ew S.S.. +
.r J'
-c.Q. After' required surface' contact is 'mettaremove Prussian Blue from top ',
l
.. }
-surface of the shim plate's'and' ~ tr surface ~bf >the' bracketi.tr Coat ;/ *,
. Vitop surface of shim plates' wit ry' film lubricant.y % l 9
(This treatment is improved by buf sng a c oth.b.k4N '
4
.o d's ':... c.
. :. J.g...pMtiW.fgQ:0... g,..Ghp qq;,. Wy.
. /
R. ' Lower the reactor vessel.to a centered position on the l shim plates.,
Check reactor vessel levelness: ;.:.?.:r.,3(.pyQ5,%;. &QWa
. X.
an;;.
,. s.; o.).T which should be within. tolerance.-
u.a S.
f (See. Step 1.) Record final results. sp%FM.;W.4.. q..m.o@ @...
- q. s..
,.q -.
T.; Check..,....the.ruactor vessel for, axis.
..y..
...,,.w..
. ;.v.. n..
a.
. orientation
..in. 'the' transit.'
~
us g'
l.[ <...l1 Recor'd. f.inal results.y.;.@gtp J:g g,,g'q
. v.d g a.,.' w...
, a q,..: ~.. x c J.nm,
,.nc,.i p.c..,. :
f
..y.
U..Take gauge readings of the. spaces on a r side o cthe support-
' V. ' Machi ne ' perinane nt'w.v$,(.WQ'('i. di,9
@(4 ' '.,e.'
' brackets for side shims id.%,
side shim plates to" mensions reco
.in Step 11 Y
less" clearance as followsi;Two *and three-loop sessels; leach si require a cold clearance of '.015" and four-loop yessels' ;,020" for;.
h.ees g of.the ; vessel support lug....,d ef'" ? Q'q.gk,j M L W.....;. - Q.?
. ys.: -
host surfaces of sid,e shim
.2.K 1
plates 1
1:contac
,/
, W.
Es, t4with'Molykote T-ype-Zidp'. film ubr ca '
above. "' j;.
hko
~
h.
esse necessa
,lat on o t
gN T
"F. M, ;; side 3hf.as'.*p1 %.MMpfp.,,-.s.
W w ;z.
..w.
.,,m
.)
.. i. c.
hoes
. t.#., 4.
Y.
Insert ' side shim plates from above' the s6pport' 7.1%..'
. m., e,.. t..~:..."d
[.9.-
0;
".,9 -
-A,',$..',g,,'yT;4g:.yM
' Z.~ Repeat Steps R. S and T.
?
a.. g4 oq t V;,., IV-2-3 VOL.!
.-s,.r-
,y
~.
s 4m..%.
I
-i_ 4i ARMS Uhlit 4c. GF l
INDEXED ~
4, -
m..,
~
~
s CONSTRUCTION OPERATION TRAVELER 351195
@ME-79-248-5500
@ EQUIPMENT NO.
@NITNO. QUANTITY (T)PAGE_l og TRAVELER NO.
TCX-RCPCRV-01 2
1 W'; e M ACTIVITY DESCRIPTION REFERENCE DRAWINGS eactor Vemmel Tnstallatinn Cao Rolnw g SPEC./ PROC./ENG. INSTR.
gLOCATION g SYSTEM MCP-1 NM containment #2 n_r.
PREPARED BY YMk DATEDDM D E PT ** - EUG-Y~/O *O QA/QC ENG ANI REVIEWED SY te DATE w. R e c t. h L A l o -z C D 4 -lo-M ANI REVIEW NA DATE QA41C l
OP. NO.
DEPT.
OPERATION,
- ENG ANI I
?
r General Notes:
(1)
All support nuts and bolts shall be tightened to " snug tight" condition.
Snuc tight shall be defined as the condition which insure that parts of the joint are brought into good contact with each other.
This can be attained by a few impacts of an impact wrench or the full effort of a man using an ordinary spud wrench.
]
Reference the appropriate.
)
Traveler for rigging and lifting require-1 ments of the reactor vessel.
(2)
Clean a I R.V.
Support Shim with acetone prior to installation.
(3)
Reference Drawings:
B
- C 2
w 1210E59--- R.V. Support Hardware QA RECORD w 1457r27--- Reactor vessel Supports OCT 02 1979
, nm, C-E 10 7 7 3-171-0 0 4----R. P.V. General riu: No.
DLES NOTED Arrangement g, 7,g g
UAEln ASGURA NC C-E 10773-171-00 5----R.P.V.
Gener 6
F.0R'IMM
~
1 M/W Install the R.P.V. Supports and Shoes per W
', L
~
Drawings 1210E59 and 1457C27.
The min.
s engagement for the guide pins is 2 ". Therefore
,j
-'the following gaps between the Support Shoe and pin top' surfaces are as follows:
4
'i
/
(
l l
t A
CONSTRUCTION OPERATION TRAVELER CONTINUATION I
TRAVELER NO.
ACTIVITY DESCRIPTION PAGG OF({
l ME-79-248-5500,
Reactor Vessel Installation PREPARED BY h --
DATE OM REVIEWED BY DATE ANI REVIEW Mb DATE I
QA/OC I
oP. NO.
Hot Leg--Shoe to pin = -2 3/8" to -7/8" Cold Leg--Shoe to pin = w3/8" to +1 1/8" W
A Ghrtn TV G
": chi.
the nJ.c f.is, Le 1 3 '4 " mn.
Pull M 77 f
The side inbed restraints to the side shim and support colu'ns prior to concrete' pour.
0" clearance must be met after the pour.
0" clearance shall be less than.001"? The' side shims are to also be modified per CI'C-4260'.' bfki" l
Weld the side shims to the R.P.V. Support per l
BRM-1457F27. QC to inspect Welds and attach applica-ble NDE Reports.
The optimum elevation to set the R.P.V.
Support shoes is as follows (based on R.P.V.
A As-Builts)-
670-----824'-5.754" l
0 158 ----824'-7.784" J
0 247 ----824'-5.755" 0
3,38 ----824'-7.777" i
The shoe bearing curface should be as level as possible, not be exceed + 1 must be within 1/8" vertIca/8".
The supports ll'y, k" radially,
" tangentially.
Record final readings on Attachment i 1. Obtain appropriate signatures.
w The Leveling screw.:shall be set as close as r'
feasibly possible to the following elevations:
l 0
E0R INED g
gU l
67 ------824'-7.004" 1580-----824'-9.034-247 -----824'-7.005" 338 -----824'-9.027" 0
4 l 'ghe leveling screws must project 1" min.--l "
'hax. above the shoe bearing surface.
i.
k
&d~r
= uM
% ma.
g u.sm /wes-catw & e+44.9% lC
@ sw i
u.s e-i i M ' '- -
j
= *
%.w
% y ot u u.
% pi m P "J' Y-h T m.Plad W43w&N
$4 4 m
'~'
' it:.......
....1
....J...
?^
~'.J~-~~
' - ~- -,
l CONSTRUCTION OPERATION TRAVELER CONTINUATION TRAVELER NO.
ACTIVITY DESCRIPTION l
ME-79-248-5500_s Reactor Vessel Installation PAGL30 2D M PREPARED BY R-DATE h #
~#~
REVIEWED BY DATE ANI REVIEW DATE i
OA/oc OP. NO.
OPERATION ENG ANI 2
M/W Lower the R.P.V.
The vessel should be level to + k" while suspended from the crane.
The QCW vessel must be orientated such that the inner and out monitor tubes straddle the 2700 azimuth.
Lower the vessel to the point of contact with the leveling screws, taking care that the vessel weight is fully imposed on all screws.
Raise any screws not in contact with the under side of the vessel support pads.
3 M/W Level the vessel to within.0005" per foot
(.007" overall) on the internals seating flange thru the access covers in the shipping fixture.
Raise the vessel and adjust the 1eveling screws as necessary.
Top of the seating flange should be set at elevation 832'-10.058".
Maintain vessel orientation using W supplied instruments in the four keyways.
Align the l
worksheet Tor /64" of containment axis.
vessel to + 1 A
leveling rea s
ckd l
{
jh h b-for field use.
4 F.E.
After obtaining proper elevation, level and orientation, recheck the calibration of all instruments to be sure no errors have be introduced. J u g/Wc h a/.5' TL 7-C-79 5
M/W Take height measurements between the under-y, QCW side of the vessel support pads and the support shoe bearing surface at four points 7g and record on' Attachment i 2.
These results must be no less than 1.90".
Also record the
' reactor cavity temperature. QC to initial and date Attachment 2.
7-ggf,- go pj 3,.
6 M/W Machine the permanent bottom shims to QCW dimensions of Operation No.,_5,,.
Tolerance'to be A
'j W W
1 I
t l
CONSTRUCTION OPERATION TRAVELER COfdTINUATION TRAVELER NO.
ACTIVITY DESCRIPTION E-79-248-5500 Reactor Vessel Installation PAGE,4g q
m D EDI%
PREPARED BY k
DATE N'#~
REVIEWED BY k
DATE ANI REVIEW
^f A DATE nA/oc OP. NO.
DEPT.
OPERATION ENG ANI 0
i.001" -
0", temp. 1 5 F.
l 7
INSV1, While bottom shims are being machined, apply p/j-R.P.V.
insulation per manufacturers supplied
,g QCM drawings and instructions..QC to write inspection report as re.1uired.
8 M/W Raise the R.P.V.
and lower the leveling screWE QCV until they are.375" +.010" above the bearing surface of the support shoen.
Install the hold'7-5-)f down nuts.
l 9
M/W Insert the bottom shim plates.
Do not allow
]
QCV the bottom shim plates to be held off.the j
bearing surface of the support shoe by the leveling screws.
E0R INE0RETlB"i DNb 10 M/W Coat the top surface of the vessel pads with prussian blue.
Lower the vessel monitoring the vessel axis to the containment axis.
T f,g 11 M/W Lift the vessel and check for surface contact QCW between the top of the shim plates and bottom %,g,gf/
of the support pads.
75% surface contact is
$7 O
required.
Snui24pisessphju$1ggregg vgpagygpodggs w a#4 84-ob etsiimgsmui sentswwwen ascessemary.
4 L.1mTg$ /
C#.2c&
gy 12 M/W After required contact area has been met, y,g),,
(pa QCy remove prussian blue from the contact surfaces 7 and coat the top surface of the shim plates Y' I
- with "molykote. type E" dry film lubricant.
rg,[,3 Buff to improve finish.
13 M/W Lower the vessel onto the shim plates.
Check the leveliness and axis orientation.
=-2===_===_____=_v_=-
v
=-
~-
~-
l l
CONSTRUCTION OPERATION TRAVELER CONTINUATION TRAVELER NO ACTIVITY DESCRIPTION g
)
l ME-79-248-5500_
Reactor Vessel Installation PAGQg DATE MNM i
PREPARED BY
~M~N DATE REVIEWED SY N
DATE ANI REVIEW oA/oc OP. NO.
DEPT.
OPERAY10N ENG ANI 4
14 MN Take gauge readings of the spaces on either 7[
QCW side of the vessel support pads and record on 2'30~77 Attachment i 3.
Also recorded cavity temperature.
gDf 15 MN Machine the side shims to the thickness D,
QCW recorded in Operation # 14. minus
.020" for 7,g 9 cold clearance.
Tolerance is + 0,
.001" l
temperature + 5 F, ggag 0
16 MN Coat the surfaces of the side shims which wilfR)
QCV be in contact with the vessel support pads g,p with "molykote type Z" dry film lubricant.
Buff to improve) finish.
j 17 MN Install the side shim in the support shoes.
l Raise the vessel to a height necessary for top installation of the shims.
Bolt the shim l
keepers onto the support shoes.-
10RJNE0RE Di l OWn 18 M/W Lower the vessel and check levelness and vessel axis for orientation with containment axis.
Record and obtain approvals on Attachment i 4 d
19 M/W verify cold clearance of.020" to 0"
.005" QCV for each side shim.
ser gy/sef s 33,go 7,g.77 fMcyS Sikp, QA review of operation traveler.T%.,
20 m
P 6'A i
Mg)
AS5emble RPV Su pped cam /s,oe.
Q each sace 4-a.,a hPuwg saati
)
{5,g rA place. Ao 44
-4.
se.F Tsa.
2" L ee,4 cJu.,
s.p.79 bug plant. 4 A powpe< e; M m M glA 4 i
I CONSTRUCTION OPERATION TRAVELER CONTINUATION TRAVELER NO.
ACTIVITY DESCRIPTION ut -79. 246. 5S00 Q ac k Ve m l 3 ukflab PAst gj p,.
b8 b4 PREPARED BY h DATE 0~
~
DATE REVIEWED BY
"^
DATE Af01 REVIEW QA/QC OP. NO.
DEPT. OPERATION ENG ANI mo it uge w
c,,s,_
a-e, e s ie u..
l2^4 wrerv w,v /w aus.u
%,x tu
- W
- HM-hers e.
o 4.s m
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a,.A i
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q ua +
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w s p y, s.
was a
4 wt c u c - 4 2s a ss
Ckq cMcAm" og
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s<
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Cs.h%4.
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er, 547
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cTW e.~, -
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&yput
,5 a <<1s M
s' nA uL w
k+c w
%e me 4
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u, a..
s,_ g I eJ 2 C eT
- I )/ "
n uw Ara w %,o, u,,,
y,,,...... _
g a
~
nn? % 91 g, s u W
u
.w, p
9 u_,a s,m p
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hf, kH
- cake.l. ww
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2.u. 5 FOR INTORMM10N 5.1
'r#")j 7.
W,a
1 1
i
/
/
CONSTRUCTION OPERATION TRAVELER CONTINUATION
/
TRAVELER NO.
ACTIVITY DESCRIPTION
/
ME M - 249 55bo 00.-deg V&
1&lc}fm PAG (OF[
['"> 4 PREPARED BY M
DATE 7~' N *'
REVIEWED BY DATE 7
ANI REVIEW DATE QA/QC OP. NO.
DEPT.
OPERATION ENG ANI b
M
- reed,
76)
Ch""f W
\\
W mev yl phi
- ' Ven h cold cla m ut. af 915 " 4* + 04 f
'I
) 49 -d saa a.a m ".
7.as 7T
'%f~
'*U b
hlg 4 4
clearamceo 6 MW
})(m%
rq ve.g A
C.l e M.A m ceo eaf-19 A wao e,Je/
c.y t4 a 8g ew~, sd ep. n M
3M* LT 7'. k.
%, 2W 6 4-9
- de 7 M 7I h %
A g n.>
meel.
lek 4el+r m w;
W""
EOR.N0m il i0; :."
r.n u
,a ~ waya w -
p3,.c loc ^h *M 1?LA 4
f th Nme J-
- 3A Ak*,
l
,g 9pec
<J.
a.lsove.
m cle m p A M "' ] W
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