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[ MEMD FOR N00 NAN: MISC 2 f9 0RANDlM FOR: VINCENT S. NOONAN, DIRECTOR, COMANCHE PEAK PRaJECT t FROM: H. H. LIVERMORE, TRT QA/0C GROUP LEADER, COMANCHE PEAK PRaJECT I
SUBJECT:
MANAGEFENT ANALYSIS COPPANY QUALITY ASSURANCE AUDIT OF j COMANCHE PEAK j THE TEXAS LITILITIES GENERATING C0ffANY (TUGCO) CONTRACTED WITH THE MANAGEMENT j ANALYSIS COMPANY (MAC) TO CONDUCT A QUALITY ASSURANCE AUDIT OF THE COMANCHE l PEAK STEAM ELECTRIC STATION. THIS AUDIT WAS CONDUCTED DURING MAY l-12, 1978.
THE AUDIT EFFORT RESULTED IN NINE BASIC NEGATIVE FINDINGS AGAINST THE COMANCHE PEAK QA PROGRAM. NLMROUS OTHER NEGATIVE FINDINGS WERE ALSO PRESENTED INDIRECTLY IN AN "0BSERVATIONS AND RECOM ENDATIONS" SECTION OF THE REPORT.
! TEXAS UTI' ITIES INFORMALLY ANSWERED WE AUDIT FINDINGS BY PEMORANDUM ON JULY lJ, I
j 1978.
i Tm TRT QA/QC GROUP WAS UNAWARE OF THIS AUDIT REPORT UNTIL IT WAS MADE
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AVAILABLE ON JUNE 1, 1985.
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IN SlM%RY, THERE ARE NO FINDINGS IN DISAGREEMENT WITH THE TRT FINDINGS; EERE ARE AREAS PRESENTED THAT WERE NOT ADDRESSED BY THE TRT; AND THERE ARE SOME f FINDINGS THAT WERE THE SAME AS THOSE FOUND BY THE TRT - FINDINGS THAT TUGC0 1
CHOSE TO TAKE NO ACTION ON.
t l FIRMLY BELIEVE THAT IF TUGC0 MANAGEMENT HAD, AT THAT TIME, TAKEN A FIRM COMITMENT TO QUALITY AND HAD ADDRESSED AND ACTED ON ALL THE IMC REC 0mENDATIONS, A GREAT NtNBER OF THE PROBLEMS WE NOW SEE AT CCNANCHE PEAK WOULD BE NONEXISTENT. ., )
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, VINCENT S. NOONAN DRAFT i .
k FOLLOWING ARE SCE OF TE NEGATIVE MC FINDINGS, RESPONSES BY TUGCO, AND Tilt j FINDINGS:
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I 1. M C FINDING: THE CURRENT PRACTICE OF AFTER-THE-FACT DESIGN CHANGE REVIEW PROVIDES SIGNIFICANT RISK OF ERROR AND IS IN KNCOPPLIANCE WITH 10 CFR 50, APPENDIX B.
TUGC0 RESPONSE: DISAGREED WITH FINDING AND PLANNED TO MAINTAIN EXISTING DESIGN CHANGE SYSTEM.
TRT-QA/QC REVIEW: THE TRT ALSO FOUND THAT DESIGN CHANGES WERE BEING REVIEWE3 BY THE ORIGINAL DESIGN ORGANIZATION AFTER TE FACT.
- 2. MC FINDING: TUGC0 QA DOES NOT REVIEW ALL PROCUREMENT DOCUMENTS AND CHANGES THERETO PRIOR TO RELEASE.
TUGC0 RESPONSE: DISAGREED WITH MC WAT EIS WAS A REQUIRENENT AND INDICATED THAT THE INCLUSION OF A QA REQUIREMENT SECTION WITH PROCUREENT DOCLEENTS WAS INADEQUATE.
! TRT QA/0C REVIEW: REGION IV (IR 50-445/84-32) FOUND SIMILAR PROBLEMS IN
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[ THE SAME AREA.
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. VINCENT S. N0ONAN DRAFT
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- 3. MC FINDING: TUGC0 WAS USING THE DESIGN CHANGE PROGRAM TO BYPASS THE
. NONCONFORMANCE SYSTEM.
4 TUGC0 RESPONSE: DISAGREED. TUGC0 WAS OF W E OPINION B AT THE DESIGN ,
CHANGE PROGRAM DID NOT BYPASS THE NONCONFORMANCE SYSTEM.
TRT-QA/QC REVIEW: THE TRT ALSO FOUND DEFICIENCIES IN NCR IMPLEMENTATION
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AND, IN SOE CASES, NCR CORRECTIVE ACTION WAS UNSATISFACTORY.
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MC FINDING: APPROXIMATELY 24% OF AUDITS SCHEDULED BY THE DALLAS STAFF l
HAVE NOT BEEN SCHEDULED. AUDITS BY TUGC0 AND BROWN & ROOT SHOULD BE COMBINED IN ONE OVERALL EFFORT.
TUGC0 RESPONSE: DISAGREED WITH FINDING. TUGC0 BELIEVED THEY COULD DEFEND l
F THE AUDIT PROGRAM AND PLANNED TO LEAVE IT AS IT WAS.
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TRT-QA/QC REVIEW: THE TRT FOUND THE PAST AUDIT SYSTEM LESS THAN ADEQUATE AND THE SYSTEM IN PLACE AT THE TIME OF THE TRT REVIEW AS QUESTIONABLE.
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! 5. MC FINDING: TIE QC INSPECTORS ARE YOUNG AND INEXPERIENCED. TERE IS T00 PtJCH RESPONSIBILITY P'_ ACED ON THE INSPECTORS IN REGARD TO DETERMINATION OF DESIGN CONFIGURATION, RESOLUTION OF PROBLEMS, AND INSPECTION PREPARATION.
M C DID NOT ADDRESS Q.C. QUALIFICATION REQUIREMENTS.
TUGC0 RESPONSE: TUGC0 DID NOT DIRECTLY ADDRESS THE FINDING. TUGCO'S i
ANSWER WAS THAT THEY WERE ALWAYS*LOOKING FOR EXPERIENCED PEOPLE. .
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L i VINCENT S. NOONAN - 'i DRAFT i
i TRT-QA/QC REVIEW: THE TRT QA/QC GROUP FOUND NUMEROUS PROBLEMS IN THE Q U INSPECTOR QUALIFICATION AND' CERTIFICATION PROGRAM.
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H. H. LIVERMORE, TRT QA/QC GROUP LEADER COMANCHE PEAK PRWECT l -
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f i MEMORANDUM FOR: Vincent S. Noonan, Director
< Comanche Peak Project FROM: Herb Livermore, TRT QA/QC Group Leader Comanche Peak Project SUBJEC1: EVALUATION OF MAC's MANAGEMENT QUALITY ASSURANCE AUDIT OF COMANCHE PEAK The Texas Utilities Generating Company (TUGCO) contracted with the Management Analysis Company (MAC) to conduct a quality assurance audit of the Comanche Peak Steam Electric Station. This audit was conducted during May 1-12, 1978.
lhe audit report was not made available to the NRC until the latter part of May 1985.
The audit was conducted by a MAC audit group consisting of a team leader and two auditors. The audit was implemented by a series of interviews with ,
utility, constructor, and architect engineer management and supervisory personnel and review of quality assurance manuals, records and work operations both at TUGC0 headquarters in Dallas and at the construction site, lhe MAC suninary of the audit indicated that recent changes in authority had not been documented in revisions to the PSAR and the Comanche Peak Quality i AssurancePlan.[Theauditalsodisclosedthat,therewereweaknessesinthe i control of design changes and nonconformance re' ports.
I b L umented in the audit Theauditeffortresultedinninefindingsthatweredy'$$b:YoftheMAC report and forwarded to TUGC0 for corrective action.
audit areas were also reviewed by the NRC Technical Review Team (TRT) QA/QC Group when they were conducting their assessment of allegations at the Comanche Peak site during 1984.~
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I i The following items paraphrases the MAC findings, the response by TUGC0 and lists the TRT QA/QC Grcup finding of the areas found deficient by MAC:
- 1. MAC Finding: The TUGC0 QA Plan and Procedures do not reflect current authoritydelegationstoBrownandRootandtoGglSandHill.
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TUGC0 Response: The deviations were incorporated into a permanent QA Plan Manual Revision on July 1, 1978.
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TRT-QA/QC Review: This area of review was not within the . scope of the TRT-QA/QC evaluation.
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! 2. MAC Finding: The current practice of after-the-fact design change review provides significant risk' of error and is in noncompliance with 10 CFR 50 Appendix 8.
TUGC0 Response: Disagreed with finding and planned to maintain existing
( design change system.
.{l l TRT-QA/QC Review: The TRT also found that design changes in some cases i were being reviewed by the original design organization after the fact.
l t 3. MAC Finding: TUGC0 QA does not review all procurement documents and changes thereto prior to release.
TUGC0 Response: Disagreed with MAC that this was a requirement and
- indicated that the inclusion of a QA requirement section with procurement documents was adequate.
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! TRT-QA/QC Review: This area of review was not within the scope of the TRT-QA/QC evaluation.
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- 4. MAC Finding: The current array of QA manuals and procedures is complex and difficult to maintain.
l TUGC0 Response: TUGC0 agreM with finding. A new plan manual was issued July 1, 1978 and the corporate QA Program Manual was under study with the goal of streamlining it.
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TRT-QA/QC Review: The TRT also found weaknesses and conflicts in quality -
assurance and quality inspection procedures. .
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- 5. MAC Finding: Records do not reflect the as-poured configuration clearly. _
TUGC0 Response: TUGC0 has taken steps to improve the visibility for the ,
i record.
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! TRT-QA/QC Review: This area of review was not within the scope of the TRT-QA/QC evaluation.
l 6. MAC Finding: Markings for in-service NDE inspections were not always distinct.
I TUGC0 Response: Agreed. Planned to have QC inspect in-service inspection markings prior to turnover.
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, TRT-QA/QC Review: This-area of review was not within the scope of the TRT-QA/QC evaluation.
[ 7. MAC Finding: TUGC0 was using the design change program to bypass the nonconformance system.
l TUGC0 Response: Disagr'eed. TUGC0 was of the opinion that the design l change program did not bypass the nonconformance system TRT-QA/QC Review: The TRT also found deficiencies in NCR implementation and in some cases NCR corrective action was unsatisfactory.
- 8. MAC Finding: The records storage facility does not have internal fire -
protection during off-duty hours.
TUGC0 Response: Agreed. An inert gas fire protection system was ,
subsequently installed.
3 TRT-QA/QC Review: The records storage facility was found to meet j applicable standards.
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I 9. MAC Finding: Approximately 24% of audits scheduled by the Dallas staff s.
j have not been scheduled. Audits by TUGC0 and Brown and Root should be .
combined in one overall effort.
TUGC0 Reponse: Diragreed with finding. TUGC0 believed they could defend the audit program and planned to leave it as it was.
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TRT-0A/0C Review: The TRT found the past audit system less than adequate h and the system.in place at the time of the TRT review as questionable.
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I f An overall evaluation of the audit indicates that MAC conducted a j comprehensive review of TUGCO's quality assurance program for Comanche Peak.
] Each of the 18 criteria of 10 CFR 50 Appendix B was addressed and nine
- specific findings were made. TUGC0 initiated corrective action for five of j the findings but did not agree with the other four findings. The four i findings for which TUGC0 did not take corrective action are discussed below
I j Finding 2: The field implementation of design changes prior to original designer review is not uncomon to nuclear construction sites recognizing that tf5 Jtility is "at risk" for the work performed. ~
- j Finding 3: TUGCO's failure to have Quality Assurance review all procurement
- documents appears to be a weakness in their program. The forthcoming adequacy i of design review for Comanche Peak will include a review of certain 19pnonarr
) procurement documents to determine if they include app-~4==+e design bases j information and applicable regulatory requirements, d
Finding 7: The design change program was being used to bypass the l nonconformance system. At the time of the TRT-QA/QC evaluation deficiencies
[ were found in NCR implementation. The weaknesses in the NCR system are identified in SSER #13 and will require response by TUGCO.
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] Finding 9: This finding by MAC was a suggestion to combine TUGC0 and Brown j and Root audit efforts. The TRT found the past aud.t system less than j adequate and the present system questionable. TUGC0 will respond to audit i deficiencies identified by the TRT in response to SSER #11.
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RALE. NUREG-0797 i Supplement No. 8 I C~'p=.u.c.ma
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Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446 Texas Utilities Generating Company, et al. U.S. Nuclear Regulatory Commission l l Office of Nuclear Reactor Regulation l . #' February 1985 ge F3.1A RRA9 f -..s, i, f
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NUREG-0797 Supplement No. 8 l l Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Linits 1 and 2 Docket Nos. 50d45 and 50-446 Texas Utilities Generating Company, et al. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation February 1985 v'" "% N.h] I i 1 e e e
5 ABSTRACT Supplement 8 to the Safety Evaluation Report for the Texas Utilities Electric Company application for a license to operate Comanche Peak Steam Electric Sta-tion, Units 1 and 2 (Docket Nos. 50-445, 50-446), located in Somervell County, Texas, has been jointly prepared by the Office of Nuclear Reactor R'egulation , and the Comanche Peak Technical Review Team of the U. S. Nuclear Regulatory Commission. This Supplement provides the results of the staff's evaluation and resolution of approximately 80 technical concerns and allegations relating to civil and structural and miscellaneous issues regarding construction and plant readiness test.ing practices at the Comanche Peak facility. Issues raised during recent Atomic Safety and Licensing Board hearings will be dealt with in future supplements to the Safety Evaluation Report.
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Comanche Peak SSER 8 iii
TABLE OF CONTENTS PaSe ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii ACRONYMS AND ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . vii
- 1. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 The Comanche Peak Technical Review Team for SER Supplement 8. . 1-2 APPENDIX K - Status of Staff Evaluation and Resolution of Technical Concerns and Allegations Relating to Civil and Structural and Miscellaneous Issues Regarding Construction and Plant Readiness Testing Practices at Comanche Peak Steam Electric Station, Units 1 and 2 . . . . . . . . . . . . . . . . . . . . . K-1 Comanche Peak SSER 8 v
ACRONYMS AND ABBREVIATIONS AA - independent assessment program allegation AB - American Bridge AB - bolt allegation ABRR - as-built reverification records A-C - Allis-Chalmers AC - concrete /rebar allegation ACI - American Concrete Institute AD - design of pipe / pipe support allegation ADS - audit discrepancy report AE - electrical allegation AE00 - Office for Analysis and Evaluation of Operational Data (NRC) AFW - auxiliary feedwater system AH - hanger allegation AI - intimidation allegation AISC - American Institute of Steel Construction ALARA- as low as reasonably achievable AM - miscellaneous allegation ANI - authorized nuclear inspector ANS - American Nuclear Society ANSI - American National Standards Institute AO - protective coating allegation AP - pipe and pipe support allegation APC - AMP Product Corporation AQ quality assurance / quality control allegation AQB QA/QC bolt allegation AQC QA/QC concrete /rebar allegation AQE QA/QC electrical allegation AQH QA/QC hanger allegation AQL - acceptable quality level AQ0 QA/QC coating allegation AQP QA/QC pipe and pipe support allegation AQW QA/QC welding allegation ARMS - Automated Records Management System ASLB - Atomic Safety and Licensing Board ASME - American Society of Mechanical Engineers - ASTM - American Society for Testing and Materials AT - acceptance test AT - test program allegation AV - vendor / generic allegation AW - welding allegation l B&PVC - Boiler & Pressure Vessel Code B&R - Brown & Root, Inc. BNL - Brookhaven National Laboratory BRHL - Brown & Root Hanger Locations BRIR - Brown & Root Inspection Report Comanche Peak SSER 8 vii
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G&H - Gibbs & Hill GAP - Government Accountability Project GDC- - general design criteria GE General Electric Corporation GED General Equivalency Diploma GHH - Gibbs & Hill hanger (isometric drawing) HFT - hot functional test NIR - hanger inspection report HP hanger package HP - high pressure i HVAC - heating, ventilation and air conditioning system HX - heat exchangers IAP - Independent Assessment Program ICC - inadequate core cooling IE - IEB Office of Inspection and Enforcement (NRC) Inspection and Enforcement Bulletin IEEE - Institute of Electrical and Electronics Engineers IM - INPO - interoffice memorandum (TVEC) Institute for Nuclear Power Operations IOM - interoffice memorandum IR - inspection report (NRC) IRN - item removal notice ITT-G - ITT Grinnell JTG - Joint Test Group (TVEC) JUMA - Joint Utility Management Assessment Group LE - left end LOCA - loss of coolant accident LP - liquid penetrant M&P - mechanical and piping MAR - maintenance action request MCC MDB - motor control center (GE) master data base MIFI - mechanical fabrication inspector MIL - material identification list (or log) MIME - Mechanical Equipment Inspector MQE Mechanical Quality Engineering MR - material requisition MRS - manufacturer's record sheet MS - main steam (line) MWDC - multiple weld data card N/A - not applicable NCR nonconformance report (TUEC) Comanche Peak SSER 8 ix
RPS - report process sheet (TUGCO) RPV - reactor pressure vessel RPVRI - reactor pressure vessel reflective insulation RRI - Resident Reactor Inspector (NRC) RV - reactor vessel RWN - room work notifications SAP - startup administration procedure SALP - Systematic Assessment of Licensee Performance (NRC) SAT - satisfactory SAVC - structural assembly verification card SER - Safety Evaluation Report (NRC) SI - safety injection SIS - Special Inspection Services SMAW - shielded metal arc welding SNM - special nuclear material SORC - Station Operations Review Committee SRIC - Senior Resident Inspector for Construction (NRC) SRP - Standard Review Plan (NRC) SRT - Special Review Team (NRC) SSE - safe shutdown earthquake ' SSER - Safety Evaluation Report Supplement SSI - _ safe shutdown impoundment SSPC - Steel Structures Painting Council SSWP - station service water pumps STE - system test engineer SWA - startup work authorization SWO - shop work order TDCR - test deficiency change request TDI - Transamerica Delaval, Inc. TDR - test deficiency report 10 CFR 50 - Title 10 Code of Federal Regulations Part 50 TI - temporary instruction TIDC - THE - Division of Technical Information and Document Control (NRC) TUEC Huclear Engineering TP - test program TPD - test procedure deviation Tr - transcript TRT - Technical Review Team (NRC) TSABC - technical services as-built coordinator TSDR - technical services design review coordinator TSI - thermolag TSMO - Technical Services Mechanical Drafting TSP - tri sodium phosphate TUEC - Texas Utilities Electric Company TUGC0 - Texas Utilities Generating Company TUSI - Texas Utilities Service, Inc.
- UCC -
University Computing Company ! USI - unresolved safety issue . I UT - ultrasonic test i UTA - University of Texas at Austin \, Comanche Peak SSER 8 xi l
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1 INTRODUCTION On July 14, 1981, the U.S. Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) (NUREG-0797) related to the application by the Texas Utilities Electric Company (TUEC) for a license to operate Comanche Peak Steam Electric Station (CPSES) Units 1 and 2. Subsequently seven supplemental Safety Evaluation Reports (SSERs) were issued by the staff. Supplement No. 7, pub-lished in January 1985, dealt with technical concerns and allegations in the electrical and instrumentation and test program areas about Comanche Peak. This report, Supplement No. 8, is the second of a series of SSERs dealing with various technical concerns and allegations about Comanche Peak. This report addresses approximately 80 technical concerns and allegations relating to civil and structural and miscellaneous issues. Appendix K to this report provides details of the staff's evaluation and findings of these technical concerns and allegations. The technical concerns and allegations about Comanche Peak were part of the regulatory issues that remained outstanding toward the completion of construc-tion of the Comanche Peak facility. The NRC's Executive Director for Opera-tions (EDO) issued a directive on March 12, 1984, establishing a program for assuring the overall coordination / integration of these issues and their reso-lution prior to the staff's licensing decision. In response to the ED0's directive, a program plan was developed and approved on June 5,1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of NRC's Region IV Office. This pro-gram plan, entitled Comanche Peak Plan for the Completion of Outstanding Regu-latory Actions, specified the critical path issues, addressed the scope of work needed, and provided a project schedule for completion. Management and coordination of all the outstanding regulatory actions for Comanche Peak are under the overall direction of Mr. Vincent S. Noonan, the NRC Comanche Peak Project Director. Mr. Noonan may be contacted by calling
- i. 301-492-7903 or by writing to the following address:
Mr. Vincent S. Noonan
- Division of Licensing l Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Copies of this Supplement are available for public inspection at the NRC's Public Document Room at 1717 H Street, NW, Washington, D.C. 20555, and the Local Public Document Room, located at the Somarvell County Public Library i On the Square, P.O. Box 1417, Glen Rose, Texas, 76043. Availability of all material cited is described on the inside front cover of this report.
l Comanche Peak SSER 8 1-1 . l -
l APPENDIX K STATUS OF STAFF EVALUATION AND RESOLUTION OF TECHNICAL CONCERNS AND ALLEGATIONS RELATING TO CIVIL AND STRUCTURAL AND MISCELLANE0US ISSUES REGARDING CONSTRUCTION AND PLANT READINESS TESTING AT COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2
TABLE OF CONTENTS P,aage
- 1. Introduction..................................................... K-1
- 2. Comanche Peak Technical Concerns and Allegations Management .
Program........................................................ K-3 2.1 Background.................................................. K-3
- 2. 2 Review Approach and Methodology............................. K-3 2.2.1 Concern and Allegation Tracking System. . . . . . . . . . . . . . . K-3 2.2.2 Review Methodology................................... K-4 2.2.3 Interviews with A11egers............................. K-5 2.3 Communication.s with TUEC.................................... K-5
- 3. Summary of Evaluations........................................... K-7 3.1 Civil and Structural Group Summary.......................... K-7 3.1.1 Scope of Concerns and A11egations.................... K-7 3.1.2 Civil and Structural Group........................... K-9 3.1.3 Findings for Civil and Structural Issues............. K-10 3.1.4 Overall Assessment and Conclusions................... K-11 3.2 Miscellaneous Group Summary................................. K-11 3.2.1 Scope of Concerns and A11egations.................... K-11 3.2.2 Miscellaneous Group.................................. K-13 3.2.3 Findings for Miscellaneous Issues.................... K-13 3.2.4 Overall Assessment and Conclusions................... K-15
- 4. Actions Required of TUEC......................................... K-16 4.1 Civil and Structural Area................................... K-16 4.1.1 Rebar Improperly Installed or 0mitted................ K-16 4.1.2 Falsification of Concrete Compression Strength Test Results................................ K-16 4.1.3 Maintenance of Air Gap Between Concrete Structures........................................... K-16 4.1.4 Seismic Design of Control Room Ceiling Elements...... K-17 4.1. 5 Unauthorized Cutting of Rebar In Fuel Handling Building.................................... K-17 4.1.6 Hollow Places in Concrete Behind Unit 2 Reactor Cavity Liner................................. K-17 Comanche Peak SSER 8 K-iii r , _ , , , _ _ _ _ , _ . _ . , . ,
- 1. . Introduction As construction of the Comanche Peak Steam Electric Station was nearing comple-tion, issues that remained to be resolved prior to the' consideration of issuance of an operating license were complex, resource intensive, and spanned more than one NRC office. To ensure the overall coordination and integration of these issues, and to ensure their resolution prior tc licensing decisions, the NRC's Executive Director for Operations (EDO) issued a memorandum on March 12, 1984, directing the NRC's Office of Nuclear Reactor Regulation to manage all necessary NRC actions leading to prompt licensing decisions, and assigning the Director, NRC's Division of Licensing, the lead responsibility for coordinating and inte-grating the related efforts of various offices within the NRC.
The principal areas needing resolution before a licensing decision on Comanche Peak can be reached include: (1) the completion and documentation of the staff's review of the Final Safety Analysis Report (FSAR); (2) those issues in contention before the NRC's Atomic Safety and Licensing Board (ASLB); (3) the completion of necessary NRC regional inspection actions; and (4) the completion and documentation of the staff's review of technical concerns and allegations regarding design and construction of the plant. Technical concerns and allegations about Comanche Peak, totalling approximately 900, have been raised mainly by the quality assurance / quality control (QA/QC) personnel working or having worked on site. Their job responsibilities involve or involved QA/QC aspects of safety-related structures, systems, and components to determine whether and to what extent such items are manufactured, purchased, stored, maintained, installed, tested, and inspected as requi.ed by project documents and procedures. Many of these allegations were made orally to NRC Region IV staff, NRC Comanche Peak Site Resident Inspectors, NRC investigators, or in letters to the NRC, as well as in testimony before the Atomic Safety and Licensing Board (ASLB). Individuals with allegations were also sponsored by the intervenor group Citizens Association for Sound Energy (CASE) and the Government Accountability Project (GAP). General allegations about poor con-struction work at Comanche Peak were also made in several newspaper articles in the Dallas / Fort Worth, Texas areas. By the end of April 1984, the staff identified approximately 400 technical con-cerns and allegations related to the construction of the Comanche Peak facility, including findings by NRC's Special Review Team. (See Section 2.1 below.) During its investigation of a concern or alTegation, the TRT identified addi-tional concerns. Interviews with allegers also yielded additional concerns. By December 1984, approximately 600 concerns and allegations had been identified. In addition, approximately 300 allegations were recently provided to the TRT by one alleger. These technical concerns and allegations were grouped by subject into the, following areas: Electrical and Instrumentation
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Civil'and Structural Comanche Peak SSER 8 K-1
- 2. Comanche Peak Technical Concerns and Allegations Management Program 2.1 Backaround Shortly after the E00's issuance of the March 12, 1984, directive, the staff found it necessary to (1) obtain current information relative to TUEC's manage-aent control of the construction, inspection, and test program and (2) obtain necessary information to establish a management plan for resolution of all out-standing licensing actions. In order to achieve these goals in an expeditious and objective manner, a Special Review Team (SRT) was formed to conduct an un-announced review of the Comanche Peak plant. The SRT consisted of eight reviewers and one team leader, all from NRC's Region II Office, and a team manager from NRC headquarters. The SRT spent over 800 manhours, from April 3 to April 13, 1984, performing this review. The SRT concluded that TUEC's programs were being sufficiently controlled to allow continued plant construction while the NRC completed its review and inspection of the Comanche Peak facility.
The SRT review also provided a basis for the development of an NRC management plan for the resolution of all outstanding licensing actions. This plan was approved on June 5, 1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of NRC's Region IV Office. The purpose of the plan was to ensure the overall coordination and integration of the outstanding regulatory actions at Comanche Peak and their satisfactory resolution prior to a licensing decision by the NRC. In accordance with the plan, a Technical Review Team (TRT) was formed to evaluate and resolve technical issues and those allegations that had been identified. On July 9,1984, the TRT began its 10-week (five 2-week sessions) ! onsite effort, including interviews of allegers and TUEC personnel, to determine the validity of the technical concerns and allegations, to evaluate their safety significance, and to assess their generic implications. The TRT consisted of about 50 technical specialists from NRC headquarters, NRC Regional Offices, and NRC consultants, who were divided into groups according to technical discipline. Each group was also assigned a group leader. I 2.2 Review Approach and Methodology i 2.2.1 Concern and Allegation Tracking System A tracking system was developed for identifying and listing each concern or allegation. These technical concerns and allegations were grouped according to their topical areas or disciplines, and were listed numerically within each group in the order that they we~e identified by the TRT. The tracking system included a description of the concern or alleqation; its status or the actions taken to resolve it; the nature of the sources of the concern or allegation (i.e., anonymous or confidential); a code for the individual who identified the concern or allegation (instead of the individual's name); the date when the concern or allegation was received by the TRT; the source document (e.g. , Comanche Peak SSER 8 K-3 m m-_ m_ --
- Based on these reviews and interviews, the TRT determined the validity of each
- technical concern or allegation and assessed its safety significance, its potential generic implications, and any indications of potential management )
breakdown. Detailed documentation of the TRT assessment and final determina- . . tions of each technical concern or allegation appear in Attachment 2 to this !~ Appendix. l 212.3 Interviews with A11egers 4 l Approximately 500 tectnical concerns and allegations regarding the construction of the Comanche Peak facility have been raised by approximately 70 allegers
.through various mechanisms. During its onsite work, the TRT interviewed 18 i individuals in person, some of whom receivec folicwup interviews by telephone.
i For ten allegers, the TRT reviewers were able to obtain the reeded information by telephone and determined that personal interviews would not be nicessary. i Three allegers contacted by the TRT declined being interviewed. Five ullagers ! could not be located during the TRT's onsite sessions because-their current , addresses and telephone numbers were not available. They have not responded l 1 to correspondence from the TRT sent to their last known addresses expressing the TRT's intention to discuss their concerns with them. Efforts to locate j these individuals included inquiries through the NRC's Office of Investigations, j NRC's Region IV staff, the telephone company and U.S. Postal Service, selected inquiries of their relatives and former co-workers, confidential examination of I the personnel files of TUEC and its contractors, and in some cases, inquiries i to the intervenor group, the Citizens Association for Sound Energy (CASE), and the Government Accountability Project (GAP).
- To the extent possible, the TRT kept a transcript for each personal interview j conducted during its onsite sessions. The names and identities of the allegers had been deleted from the transcripts, as well as from other pertinent reference
! or source documents, before TRT reviewers were given any portions of these
- documents for review and follow-up. During the TRT's onsite work, tt.e original transcripts were kept in a locked file in the TRT Project Director's office.
The distribution of these transcripts within the NRC, and even within the TRT, i was limited and controlled. l ' l Subsequent to its onsite work, and at the completion of its evaluation, the TRT
- attempted to contact each alleger to discuss the TRT's findings regarding their original concerns, and to obtain additional comments from them, if any. Thirty i
' allegers have received such followup interviews. A total of 19 allegers could not be located. Some of these individuals had received initial TRT interviews . but had since left the area. Three allegers declined to have further contacts with the TRT. The TRT is in the process of contacting the remaining allegers ! for followup interviews. The outcome of followup interviews conducted through j December 1984, is briefly discussed in the individual SSER sections in Attach- ! ment 2. Transcripts were kept for all followup interviews conducted either by
- telephone or in person.
2.3 Communications with TUEC Whenever the TRT reviewers encountered problems during their evaluations, the l TRT Project Director and/or his designee resolved them through discussions with ! TUEC management onsite. There were also frequent staff-level contacts between TRT members and TUEC personnel during the TRT's onsite activities. In keeping )
- Comanche Peak SSER 8 K-5 l .
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, - --_... __- - _ _ _ . _ . . . , . . - _ _ , , - _ - - ~ . _ -._ _ _.___ -
- 3. Summary of Evaluations 3.1 Civil and Structural (C&S) Group Summary 3.1.1 Scope of Concerns and Allegations The concerns and allegations in the C&S discipline involved most aspects of reinforced concrete construction and testing. These allegations and concerns relate to (1) design deficiencies, (2) testing or inspection irregularities, (3) incorrect construction practices, (4) inadequate repairs, (5) uncorrected, unsafe conditions in the completed structures, and (6) premature structural loading. The total of 57 concerns and allegations were grouped by subject into the following 17 categories:
Category Characterization of ho. Subject Concerns and Allegations 1 Inadequate Materials Used in Rejected aggregate was incor-Concrete porated in the basemat of Unit I reactor; unauthorized quanti-ties of water added to concrete used in basemat; rejected con-crete placed in turbine genera-tor building; concrete with excessive slump placed in con-tainment walls; concrete rejected for being over speci-fication limit on time to dis-charge was placed in the Circu-lating Water Intake Structure. 2 " Bad Concrete Work" and " Bad concrete work" and " slop-
" Sloppy" Placement of py" placement of concrete; place-Concrete ment of " soupy" concrete in a slab in the Auxiliary Building j in the summer of 1976.
- 3 Placement of Concrete Placement of concrete during During Poor Weather Conditions rainstorm and without approval
! by QC personnel and during or immediately before freezing weather; some field-cured cylinders and standard-cured cylinders failed specification requirements for concrete strength, and the Schmidt rebound hammer test was j misapplied. l
, Comanche Peak SSER 8 K-7 l
l
Category Characterization of No. Subject Concerns and Allegations 11 Poor Workmanship in Use of Poor workmanship in use of Rotofoam rotofoam as a temporary spacer during construction to maintain required seismic gap between Category I concrete structures. 12 Concrete Construction A spillway pillar, span, or
. Deficiencies column was erected 75 to 80 degrees offset.
13 Concrete Cracks At Bottom Detrimental cracks in concrete of Reactor Vessel pad at bottom of reactor vessel. 14 Control Room Area Deficiencies The field run conduit, drywall, and lighting fixtures installed above ceiling panels in the control room are classified as nonseismic and are supported only by wires, and may fall as a result of a seismic event. 15 Unauthorized Cutting of Rebar Undocumented and unauthorized holes were drilled through rebar. 16 Excavation Overbreak/ Overexcavation and improper Seismic Response fill under Unit 1 Containment Building could invalidate expected seismic response of the foundation due to change in properties resulting from removal of in-situ material. 17 Improper Concrete Sampling Personnel produced incorrect f readings on concrete batch
+
plant scales by leaning on wires connecting the weighing hoppers to'the scales. 3.1.2 Civil and Structural (C&S) Group The Civil and Structural Group consisted of three NRC employees and four consul-tants, all of whom are civil and structural engineers, with a combined total of 137 years of experience in general design and in nuclear and non-nuclear heavy construction work. These reviewers were selected for their technical expertise and experience in design, construction, quality assurance, and ability to detect discrepancies in construction records. a Comanche Peak SSER 8 K-9 m -e
strength of the concrete placed in Category I concrete structures. (See Attachment 2, C&S Category 8.) Another issue that was not substantiated and whose safety significance could not yet be determined concerned the unauthorized cutting of rebar in the Fuel Handling Building. The C&S Group found that if certain holes were drilled to the depth alleged, rebar would have been cut without authorization. (See Attachment 2, C&S Category 15. ) The C&S Group founc' that 'the allegation concerning hollow places in concrete behind the stainless steel liner of the Unit 2 Reactor Cavity is true; the . hollow places are currently undergoing repairs. The repairs and the repair documentation must be inspected, reviewed, and approved by the NRC before the TRT can determine that this issue has been adequately resolved. (See Attach-ment 2, C&S Category 4.) The C&S Group could not substantiate the concerns raised by the remaining alle-gations and concluded that these concerns have no structural safety significance. However, the results of the evaluations for Categories 1,3,4,5,6,7,8,9, 10, 15, and 17 are being further assessed by the QA/QC Group as part of its overall programmatic review. (See Attachment 2, C&S Categories 1, 3, 4, 5, 6, 7, 8, 9, 10, 15 and 17.) 3.1.4 Overall Assessment and Conclusions During its evaluations, the TRT reviewed pertinent construction records (e.g., concrete placement packages, NCRs, concrete test results), structural design drawings and calculations, specifications (e.g., for concrete reinforcing steel), interviewed craft and TVEC personnel and conducted plant inspections. This documentation, to the extent reviewed by the TRT, was judged to be ade-quate and consistent with applicable FSAR commitments, except for the deficiencies identified in the SSER sections in Attachment 2. Therefore, the TRT concludes that the civil and structural construction within the scope of the TRT C&S group review effort was adequate and was, for the most part, well documented. 4 <e issues in the civil and structural area still require further action. One case involving reinforcing steel omitted from the reactor cavity wall, and another case of alleged unauthorized drilling of reinforcing steel, require further documentation. TUEC must also test concrete in place to evaluate an allegation concerning falsified concrete strength tests. In addition, TVEC must conduct analyses and inspections to determine whether the separation between buildings is adequate to provide acceptable performance in o earthquake. Finally, there must be a seismic analysis of the suspended ceiling, lighting fixture and nonsafety-related conduit in the control room to demonstrate design adequacy of the ceiling elements. The potential safety implications of this issue for nonseismic structures, systems, and components in other parts of the plant must also be evaluated. 3.2 Miscellaneous Group Summary 3.2.1 Scope of Concerns and Allegations . The allegations with a Miscellaneous designation covered a wide variety of topics and involved both administrative and construction activities, some Comanche Peak SSER 8 K-11
Category No. Subject Characterization of Concerris and Allegations 13 Tube to Base Tube steel was cut at the wrong angle and welded Plate Weldments to a baseplate, leaving a large gap between the tube and baseplate. - 14 NRC Form-3 NRC Form-3 was posted at an insufficient number Posting of site locations. 15 Drug Abuse Drug use and abuse was widespread and management did not give proper attention to the alleged problem. 16 HVAC Heating, ventilating, and air conditioning system (HVAC) supports for seismic loads were not ana-lyzed; HVAC components and supports inside con-tainment were not properly considered as missiles; HVAC failure during a postulated accident would allow temperatures to rise to an unacceptable level inside containment. 17 Reactor Vessel Damage occurred to upper internals of the reactor Internals vessel. 18 Polar Crane Internal wires were broken in the polar crane Cables festooned cables. 19 Radwaste System Workers habitually urinated on stainless steel Contamination pipe. 20 Instructions to Inadequate rigging and handling instructions were Craft Personnel provided to craft personnel. 3.2.2 Miscellaneous Group The members of the Miscellaneous Group were assembled based on their technical expertise, capabilities, and experience in engineering design, quality assur-ance and document control, inspection, construction, and regulatory activities. The group included five members from NRC's Region IV office, with expertise in various technical disciplines, and three consultants. Collectively, the group possessed experience in excess of 50 years in the nuclear power industry and its regulation. 3.2.3 Findings for Miscellaneous Issues Fourteen of the 24 allegations (grouped into 20 categories) pertained to systems and components classified as nonnuclear safety (NNS) in the Comanche Peak Steam Electric Station (CPSES) Final Safety Analysis Report (FSAR). These allegations (AM-3, 4, 5, 6, 9, 15, 16, 17, 22, 24, 25, and 30) were also not listed as quality assurance (i.e., safety-related) items in Table 17A of Volume XIV of the FSAR. Accordingly, 10 CFR Part 50 Appendix B quality assurance requirements would not apply to these systems and components except Comanche Peak SSER 8 K-13
During further investigation of the allegation that shims for the rail support system of the polar crane had been altered during installation, gaps, which may have been excessive, were observed between the crane girder and the girder sup-port bracket. Detailed specifications addressing the gap tolerance in the girder seat connections did not exist; however, Gibbs & Hill indicated in a November 28, 1977 letter (GHF-2207) that seated connections do not require shimming, since the area in bearing is at least the width of the bottom flange of the crane girder. Contrary to this assumption, nine girders were observed to have gaps which extended under the bottom flange that reduced the bearing surface to less than the 20-inch flange width stated in the letter. The TRT also observed conditions which indicated that the crane rail may still be moving in a circumferential direction, that three rail-to-rail ground wires were broken, that two shims have partially worked out from under the rail, and that two Cad-welds were broken. (See Attachment 2, Miscellaneous Category 11.) The TRT found that 21 of 24 allegations (that is, all except AM-3, 15 and 16) were either unfounded or involved nonsafety-related issues, or the deficiency was identified by TUEC's quality assurance / quality control program and correc-tive actions had been completed that were acceptable to the TRT. 3.2.4 Overall Assessment and Conclusions The TRT found that 9 of the 24 allegations were substantiated, were potentially safety significant, and had generic implications. However, actions taken because of NRC Bulletins, inspections, and TUEC audits / evaluations corrected all but two problems. Therefore, the TRT concludes that 21 of 24 allegations had neither safety significance nor generic implications. The two problems for which TUEC will have to complete actions and address issues are Miscellaneous Category 2, the gap between the reactor pressure vessel reflective insulation and the biological shield wall, and Miscellaneous Category 11, improper shimming and installation of the polar crane rail support system. (See Section 4.2.) Once these actions are satisfactorily completed by TUEC and are reviewed and accepted by the NRC, a finding can then be made that no outstanding issues raised by the miscellaneous concerns and allegations remain that would preclude licensing of CPSES Unit 1. Comanche Peak SSER 8 K-15
4.1.4 Seismic Design of Control Room Ceiling Elements (See Attachment 2, C&S Category 14)
- Provide the results of seismic analysis which demonstrate that the non-seismic items in the control room (other than the sloping suspended dry-wall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- Provide an evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceil-
.; ing (nonseismic item with modification) which accounts for pertinent floor ;
response characteristics of the systems.
- Verify that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy appli-cable design requirements.
- Provide the results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit 2 inches or less in diameter.
- Provide the results of an analysis which demonstrate that the foregoing problems are not applicable to other category II and nonseismic structures, systems, and components elsewhere in the plant.
4.1.5 Unauthorized Cutting of Rebar in the Fuel Handling Building (See Attach-ment 2, C&S Category 15)
- Provide information to demonstrate that only the No. 18 reinforcing steel in the first layer of the floor slab at the 810-ft, 6-inch elevation of the Fuel Handling Building was cut during installation of the trolley process aisle rails, or
- Provide design calculations to demonstrate that structural integrity is maintained if the No. 18 reinforcing steel on both the first and third layers of the floor slab was cut.
4.1.6 Hollow Places in Concrete Behind Unit 2 Reactor Cavity Liner (See Attachment 2, C&S Category 4)
- Provide details of the successful completion of the repairs to the hollow placas in concrete behind the Unit 2 reactor cavity liner.
4.2 Miscellaneous Area 4.2.1 Gap Between Reactor Pressure Vessel Reflective Insulation (RPVRI) and the Biological Shield Wall (See Attachment 2, Miscellaneous Category 2)
- Review the procedures for approval of design changes to non-nuclear safety-related equipment, such as the RPVRI, and make revisions as necessary to assure that such design changes do not adversely affect safety-related systems. -
Comanche Peak SSER 8 K-17
7 I ATTACHMENT 1 LISTING OF TECHNICAL CONCERNS AND ALLEGATIONS I. Civil and Structural Allegation - Number Characterization Category Page Number AQC-1 Concrete air entrainment test records 8 K-59 were falsified. - AQC-2 Concrete laboratory test rccords were 8 K-59 falsified. AQC-3 Concrete aggregate tests were falsified. 8 , K-59 AQC-4 Equipment required for aggregate testing 10 K-71 had not been used. AqC-5 Improper methods were used to' dry coarse 10 K-71 aggregate for sieve analysis. AQC-6 Some of the Unit 1 Containment Building 10 K-71 basemat concrete was placed without required testing. AQC-7 Concrete compressive strength test 8 K-59 results were falsified. AQC-8 Concrete compressive strength test spect- 10 K-71 4 mens were loaded at an excessive rate. AQC-9 Recertification examinations for 9 K-67 R. W. Hunt inspectors were given open book and examinations were given with answers supplied. AC-10 Concrete repair following removal of a 7 K-57 Hilti bolt was improper. AQC-11 Acceptable concrete test cylinders were 10 K-71 used to represent concrete placements other than those for which the samples were made. AQC-12 Reinforcing steel (rebar) was installed in 6 K-49 the Containment Building without quality control (QC) inspection. K-19
i l I. Civil and Structural (Continued) Allegation Number Characterization Category Page Number AC-25 Voids existed in the concrete wall behind 4 K-39 the Unit I reactor cavity stainless steel liner. AC-26 Equipment was set on grout before the 5 K-45 grc. properly gained its requ. ired strer;th through curing. AC-27 Rejected and improper material was used 1 K-27 in concrete batches. (See AQC-16, AC-19, AC-20,AC-21,AC-47.) AC-28 Fresh concrete was placed on top of 4 K-39 crumbling concrete during construction of a spillway. AC-29 A pillar, span, or column associated with 12 K-79 a spillway was erected 75 to 80 degrees offset. AC-30 Recer was omitted from a portion of the 6 K-49 Safeguards Building , AC-31 Richmond Insert anchor bolt inserts were 5 K-45 installed in Unit 1 at angles not per-pendicular to the concrete surface. J AC-32 A 20-ft by 20-ft area of honeycombed con- 4 K-39 crete in the Unit 1 Auxiliary Building was inadequately repaired. AC-33 Cracks exist in the Unit 1 concrete 4 K-39 basemat and in the floors of other plant buildings. AC-34 Concrete voids could be detected in 4 K-39 building walls by tapping with a hammer and listening for a hollow sound. AC-35 Concrete was placed in the Safeguards 3 K-33 Building basemat and the lowest level floor of the Unit 1 Containment Building during or just before freezing weather. AC-36 Trash was placed in a form and then 5 K-45 covered with concrete.
. K-21
I. Civil and Structural (Continued) Allegation Number Characterization Category Page Number AC-49 Rebar was installed upside down in a build- 6 K-49 ing near the Unit 2 containment structure. AC-50 " Soupy" concrete was placed in a slab in 2 K-31 the Auxiliary Building during_the summer of 1976. AQC-51 Cadweld tensile test results were 8 K-59 recorded during the spring and summer of 1976 without the tests having been performed. AC-52 Several examples of field-cured cylin- 3 K-33 ers and standard-cured cylinders failed specification requirements. The Schmidt rebound hammer test was then misapplied to resolve these test failures. AQ-64 Overexcavation and improper fill under 16 K-93 the Unit 1 containment structure could invalidate the expected seismic response of the foundation due to changes in properties from the removal of in-situ material. Miscellaneous (Issues from this allegation were OAM- addressed by the Electrical Group [AE-50 and AE-51]; the Mechanical and Piping Group [AP-24 AP-25, AP-26,AP-27,AP-28,AQW-69,AQW-71];
,(
L and the QA/QC Group [AQ-111].) ggP AM-2 Nuclear fuel was received onsite 1 K-97 before the NRC issued a special materials license. AM-3 During hot functional testing, 2 K-99 expansion caused the reactor pressure vessel reflective insulation to touch the biological shield wall. AM-4 The Preliminary Safety Analysis Report, 3 K-103 Sections 10.2-11 and 10.2-12, contained errors. K-23
II. Miscellaneous (Continued) Allegation Number Characterization Category Page Number AM-18 The tube steel used to fabricate 13 K-127 supports in the Unit 1 safeguards "796 yard tunnel" was cut at the wrong angle, resulting in excessive gaps for the weld joints between the tube steel and baseplates. AM-19 The posting requirements for NRC 14 K-131 Form 3 were not met from 1977-1982. AM-20 Material false statements were made by plant management to the Atomic Safety and Licensing Board. (This allegation was transferred to the NRC Office of Investigations for followup.) AM-21 There was widespread drug abuse 15 K-133 at Comanche Peak, and management did not give proper attention to this problem. AM-22 TUEC has not analyzed the heating, 16 K-137 ventilating, and air conditioning system (HVAC) supports for seismic loads. HVAC components and sup-ports inside containment were not properly considered as missiles. HVAC failure during a postulated
. accident would allow temperatures to rise to an unacceptable level inside containment.
AM-23(a) A craft person stated that he had 20 K-147 not received instructions about how to rig and handle a large motor-operated valve. AM-23(b) The Unit i reactor pressure vessel 2 K-99 is located 3/16 inch west of the north-south centerline through the containment building. AM-24 15-foot by 2 -inch stainless steel 17 K-139 bars inside the Unit I reactor vessel upper internals were damaged and then repaired without proper documentation. 4 K-25 ,
ATTACHMENT 2 ASSESSMENT OF INDIVIDUAL TECHNICAL CONCERNS AND ALLEGATIONS IN CIVIL AND STRUCTURAL AND MISCELLANEOUS AREAS
- 1. Allegation Category: Civil and Structural 1, Inadequate Materials Used in Concrete
- 2. Allegation Number: AQC-16, AC-19, AC-20, AC-21, AC-27 and AC-47
- 3. Characterization: It is alleged that t'he following violations of specifi-cations occurred at various times:
- a. Rejected aggregate was incorporated in the basemat of the Unit 1 reactor (AQC-16).
- b. Truck drivers added unauthorized quantities of water to concrete used in the basemat (AC-19).
- c. Rejected concrete was placed in the turbine generator building (AC-20). .
- d. Concrete with excessive slump was placed in containment walls (AC-21).
- e. Some concrete was placed in the Circulating Water Intake Structure after the concrete was rejected for being over specification limit on time to discharge (AC-47).
AC-27 contained no new allegations; it merely reiterated those already made. Allegations AC-19, AC-20, and AC-21 were investigated by Region IV and documented in inspection report 79-09, which was reviewed by the NRC Tech-nical Review Team (TRT) as a step in its own assessment of the allegations.
- 4. Assessment of Safety Significance: Allegations AC-19, AC-20, and AC-21 appeared in a newspaper article. The identify of the alleger of AC-19 was not disclosed and therefore could not be contacted. Allegations AQC-16 and AC-47 were judged as having sufficient clarity for technical resolu-tion without initial contact between the TRT and the allegers.
- a. The TRT cannot determine whether or not the allegation that " rejected" aggregate was used is valid (AQC-16). The only item in the record was Deficiency and Disposition Report C-446 (December 9, 1976), which stated that an untested pile of aggregate, rather than an unacceptable pile, was used. Therefore, the acceptability of the aggregate is unknown. .The alleger also stated that the equipment operator scraped aggregate off the floor of the storage area and dumped it on the conveyer belt so that it bypassed testing. The consequences of this alleged action may be evaluated by the effect on the properties of fresh and hardened concrete. The purpose of controlling aggregate grading is to maintain concrete of uniform workability and strength.
A TRT examination of the concrete basemat placement record packages revealed that workability and strength were satisfactory throughout the placement. Less than 3 percent of the concrete was rejected for improper slump, and all concrete tested met the specifications for K-27
Thus, placement of a batch with a 4 -inch slump was permitted as long as the average of all batches or the most recent 10 batches did not exceed 4 inches and individual batches did not exceed 5 inches. A single high slump batch, provided the slump does not exceed 5 inches, cannot constitute a violation of specifications.
- e. The TRT reviewed 51 (37%) concrete packages, out of the 140 concrete packages for the Circulating Water Intake Structure which is a non-safety-related structure (FSAR Vol. IV, Sec. 3.2) (AC-47). Of the 51 reviewed, 13 batches of concrete were rejected, 9 for test failure (air, slump, temperature), and 4 for being over the specification limit on time to discharge. It was noted on the batch ticket of each rejected batch of concrete where the con' crete was dumped. None of the rejected batches was placed in the circulating water intake; they were placed in temporary slabs which the contractor was placing at the time.
- 5. Conclusion and Staff Positions: The allegations were found to have no structural safety significance.
Based on a review of pertinent documentation, test results, and the concrete placement packages, the TRT concludes that if nonconforming aggregate was used in the basemat of the Unit 1 reactor, it did not adversely affect its concrete properties. The only indication of water addition found by the TRT was within the stipulated limits, thus ensuring that there was no adverse effect on the concrete. However, the absence of laboratory signatures on batch tickets represents a failure to follow QA/QC program requirements. The results of the evaluation pertaining t the lack of laboratory signatures will be further assessed as part of the overall programmatic review concerning procedures addressed under QA/QC
, Category 6 "QC Inspection." Therefore, the final acceptability of thi evaluation will be predicated on the satisfactory results of the program-matic review of this subject. In its examination of 65 random samples of concrete placements in the turbine building, which is a nonsafety-related structure, the TRT found no evidence of irregularities. Batch placements were within tolerances specified by G&H and the TRT found no documenta-tion that these slump requirements had been violated. ;The pl eement of a single batch of concrete with a 4%-inch slump does notl constitute a violation of specifications. Thebatchticketsstatejthatnoneofthe rejected concrete batches was placed in the circulating water intake; therefore, the TRT concludes that the allegation is without foundation.
The alleger of AC-19 was not identified so that the TRT could not conduct a closing interview. The TRT is attempting to contact the individual who made allegation AC-21. The individual who made allegation AQC-16 did not wish to meet any further with the TRT and will be informed of the pertinent TR1 findings by letter. The individual who made allegation AC-20 declined to be interviewec. by the TRT and will also be informed of the pertinent TRT findings by letter. The alleger of AC-47 could not be located for a closing interview.
- 6. Actions Required: None.
K-29
- 1. Allegation Category: Civil and Structural 2, Concrete Placements
- 2. Allegation Number: AC-22, AC-23 and AC-50
- 3. Characterization: It is alleged that " bad concrete work" and " sloppy" placement of concrete occurred at the Comanche Peak Steam Electric Station (CPSES) (AC-22, AC-23). It is also alleged that " soupy" concrete was placed in a slab in the Auxiliary Building in the summer of 1976 (AC-50).
- 4. Assessment of Safety Significance: The individual making allegations -
AC-22 and AC-23 was in'.erviewed by the NRC Technical Review Team (TRT). Allegation AC-50 was judged as having sufficient clarity for technical resolution without initial contact between the TRT and the alleger. In testimony at an Atomic Safety and Licensing Board (ASLB) hearing, the first alleger did not identify a particular structure or concrete place-ment that had " bad concrete work" or that exhibited " sloppy" placement cf concrete. To adequately encompass the concerns raised in this allegation, the TRT reviewed random samples of concrete placement packages from three safety-related buildings to determine if the allegations were valid. A review of 14 packages from the Auxiliary Building and 3 packages each from the Unit 1 Safeguards Building, the Unit 1 Containment Building (exterior), and the Unit 2 Containment Building (exterior) revealed that the quality control (QC) inspector accepted the forms and reinforcing steel placement prior to each concrete' placement. Of the 23 placement packages reviewed,10 had nonconformance reports (NCRs) related to con-crete placement. One package had four NCRs, two other packages had two NCRs each, and the remaining seven each had one NCR. Seven NCRs were resolved with the designation "use-as-is," seven with " repair," and one with " reject." The seven placements indicating " repair" were for concrete honeycombing; the one indicating " reject" was for the removal of concrete from a small pad. In addition to a records review, the TRT performed a walkthrough inspection of the safety-related buildings. The defects treated in these NCRs are visible from the surface and were examined by the TRT. The TRT concluded that there was no degradation in' quality in any of the observable concrete surfaces. The TRT also interviewed two QC inspectors at Comanche Peak who were concrete placement inspectors on some of the concrete placements in the Auxiliary Building reviewed by the TRT. Both QC Inspectors stated that they were not cognizant of any " bad concrete work" and/or " sloppy" place-ment of concrete at CPSES. They stated that all personnel with construc-tion and concrete placement responsibilities would meet prior to each placement to resolve any potential problems. They stated that for the concrete placements they were involved with, the work was done in accord-ance with project procedures and other pertinent requirements. They also stated that niacement crews cooperated with requests from QC personnel. The individual who made the al. legations discussed above was contacted by the TRT to inform him of the TRT's finding. The alleger expressed his satisfaction with respect to the TRT's disposition of his allegations. K-31
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- 1. Allegation Category: Civil and Structural 3, Poor Weather Conditions
- 2. Allegation Number: AC-24, AC-35 and AC-52
- 3. Characterization: It is alleged that the placement of some concrete took place under the following adverse weather conditions: (a) during a rain-storm and without the approval of quality control (QC) personnel (AC-24) and (b) during or immediately before freezing weather (AC-35). It is further alleged that (c) there are several examples of field-cured cylinders which failed specification requirements, that some standard-cured cylinders failed specification requirements, and that the Schmidt rebound hammer test was misapplied in resolving problems created by these deficiencies (AC-52).
- 4. Assessment of Safety Significance: Allegation AC-24 was th'e subject of testimony given by Region IV inspectors, but the identity of the alleger, was not revealed. The TRT attempted to determine the alleger's identity but could find no record of it. Allegation AC-35 was judged as having sufficient clarity for technical resolution without initial contact betweer the NRC Technical Review Team (TRT) and the alleger. The TRT interviewed the alleger of AC-52.
- a. In assessing the allegation concerning placement during a rainstorm (AC-24), the TRT examined concrete placement package 101-8805-013 for a placement on the dome of the Unit 1 Containment Building. This package indicated that the final batch of concrete placed :n the evening of January 18, 1979, was batched at 5:59 p.m.; that only about 300 of the required 450 cubic yards had been placed; that the crew abandoned the placement in a heavy rain at 7:30 or 8:00 p.m. , leaving a gap with a 30-foot radius in the middle of the placement; that concrete batching started again at 8:00 a.m. on January 19; and that the lift was topped out at 12:21 p.m. There is no account of any irregularity during the shutdown. However, the craft personnel general foreman for the placement reported that the attempt to cover the partially completed concrete with plastic to protect it from the rain was not completely successful; that about a half cubic yard of concrete was washed out before the crew got the situation under control; that at about 10 p.m. he went to the batch plant, which was now empty because of the hour, dry-batched a half cubic yard to the correct proportions, mixed it in two batches in the concrete labora-tory mixer, and placed it on the dome. At this time, all quality control personnel had gone home and were not available to approve or oversee the operation. This sequence of events was not refuted by the NRC Region IV investigation (inspection report 79-11) of this incident and is apparently correct. The TRT interviewed the author of the Region IV inspection report.
The action constitutes a violation of 10 CFR 50, Appendix B, Criterion X and indicates a partial breakdown of the quality control system. Following the completion of the dome, and after learning of the allegation of a violation, Brown & Root engaged Muenow and Associates to make an ultrasonic investigation of the portions of the K-33
suspect concrete and the acceptable concrete at an age of 4 months. The results are recorded in HCP reports 10664 and 10849, which were inspected by the TRT. For both series of tests rebound num-bers ranged fron 39 to 46. The concrete on the edge adjacent to the dowels, which was difficult to protect, is acceptable for the following reasons: (1) it was not exposed to freezing tempera-tures for 6 days following its placement; (2) concrete at that age should not be damaged by freezing; and, (3) Schmidt hammer readings were the same on suspect concrete as on well protected concrete. (2) Placement package 101-2808-001 reports on the concrete in the Unit 1 containment structure, which was placed on December 30, 1976. On the evening following the placement, the ambient temperature dropped below 20*F. The records showed a concrete surface temperature as low as 42*F during the first day and no surface temperatures below 50*F on subsequent days, in spite of the fact that ambient temperatures as low as 12*F were measured. The protection, as indicated by the records, complied with specifications. However, the allegation was triggered by an event detailed in Brown & Root (B&R) interoffice memo IM 7700. During the first evening, a TUEC QC inspector measured a surface temperature of 21*F. The B&R QC inspector noted that the TUEC inspector used an uncalibrated thermometer with a large range and took the reading in such a manner that the thermometer was not protected from the air so that in the B&R QC inspector's opinion, the TUEC inspector was measuring ambient temperature instead of the concrete surface temperature. Although the two discussed the adequacy of the technique, and a picture was taken of the technique, the records did not indicate that the matter l was ever resolved. To evaluate the condition of concrete alleged to have been exposed to freezing temperatures, the R.W. Hunt Co. ran Schmidt hammer tests on the suspect concrete and on concrete whose integrity was not in doubt. The results are in HCP report 22014, which was examined by the TRT. Rebound numbers for
- i. suspect areas ranged from 25 to 35, and in sound areas from 27 f to 36. The differences are not significant.
- c. The allegation concerning field-cured test cylinders, standard-cured l test cylinders, and Schmidt rebound hammer tests (AC-52), is
! contained in the attachments to a letter, dated September 20, 1984, to Thomas Ippolito, NRC, from Mrs. Juanita Ellis, President of the Citizens Association for Sound Energy (CASE). The allegation states, l " Based on a review of documents attached and already in the record, ! it is apparent that the quality and compressive strength of the concrete at Comanche Peak is indeterminate at best, and, in some cases appears to be deficient." This observation is presumably supported by Attachment D to the letter, which lists 36 test cylinders in 18 placements with laboratory-cured strengths below 4000 l psi. The alleger was interviewed, and he stated that he was under the impression that the concrete was designed for a strength of 4000 psi. The TRT reviewed the, records and found that all the cited K-35
= .-
The allegation questioned the use of the Schmidt rebound hammer for qualifying sections of concrete in which field-cured test cylinders failed to meet specifications. The above discussion makes the issue moot except for the single cylinder in a slab of the Safeguards Building. The use of the Schmidt rebound hammer has general acceptability and is specifically permitted by the Comanche Peak construction specifications as an aid in evaluating concrete strength in place, as discussed below. ASTM C-805 states, "The rebound number determined by this method may be used to assess the uniforaity of concrete in situ, to delineate zones or regions of poor quality...." Paragraph 7.3.e of the p,roject specifications states, " Evaluation of test results shall be in accordance with Section 17.1, 17.2, and 17.3 of ACI 301." Section 17.3.1 of ACI 301 states, " Testing by impact (Schmidt) hammer, soniscope, or other nondestructive device may be permitted by the architect / engineer to determine relative strengths at various locations in the structure as an aid in evaluating con-crete strength in place or for selecting areas to be cored. Such tests, unless properly calibrated and correlated with other test data, shall not be used as a basis for acceptance or rejection." Hammer results are normally not permitted as a substitute for laboratory-cured test cylinders, which form the basis for acceptance of the concrete. They may be used to judge the ade-quacy of protection or to determine when a portion of a structure may be safely loaded. The ACI Building Code and the Comanche Peak specifications do not provide for the rejection of concrete on the basis of low-strength, field-cured cylinders. They merely require that protection be improved and that critical elements be cured for a longer period of time before being loaded. With the exception noted above, the low field-cured strengths were not in critical elemen . The statement in Attachment D that all retesting which had been promised had not been carried out is a quality assurance matter, not a safety problem, and it will be investigated by the TRT QA/QC group The statement that Schmidt hammer tests were not conducted on sectio of concrete when both field-cured and laboratory-cured cylinders were below 4000 psi does not appear to be pertinent since there were no sections cited where both field and laboratory results were below the design strength.
- 5. Conclusion and Staff Positions: Although these allegations are true, they do not tave structural safety significance.
(a) The Unit 1 dome was proved sound both by ultrasonic testing and by structural integrity testing. (b) Sections of concrete alleged to have been exposed to freezing temperature at an early age were shown by in place strength tests to have substantially the same strength as concrete whose protection was not in doubt. (c) The field-cured test cylinders demonstrated adequate protection for the type of concrete placed, with the exception of one slab in the Safeguards Building, which was shown by Schmidt hammer testing to be adequate. - l K-37
b 4 Y
- 1. Allegation Category: Civil-and Structural 4, Concrete Voids and Cracking
- 2. Allegation Number: AC-25,iAC-28, AC-32, AC-33 and AC-34
- 3. Characterization: It is alleged that the following concrete deficiencies occurred at Comanche Peak Steam Electric Station (CPSES):
- a. Hollow places existed in concrete behind the stainless steel liner of the Unit 1 reactor cavity (AC-25).
- b. Fresh concrete was placed on top of crumbling concrete during the construction of the spillway (AC-28).
- c. The repair of a 20-foot x 20-foot honeycombed area located in the Unit 1 Auxiliary Building was inadequate (AC-32).
- d. Cracks existed in the concrete reactor cavity wall of Unit 1 and in floor slabs in the plant buildings (AC-33).
- e. There are numerous concrete voids in building walls that can be located by tapping the walls with a hammer and listening for a hollow sound (AC-34).
Allegation AC-25 was investigated by Region IV and documented in inspection reports 80-08 and 80-11, which were reviewed by the TRT as a step in its own assessment of the allegation. _ In addition to these allegations, the Region IV resident inspector requested that the TRT review the following possible reportable design i deficiencies involving concrete placing problems.
- f. Reportable Design Deficiency Concerns:
4 (1) A void was identified in the Unit 1 Reactor Building Steam Generator Compartment Wall. I (2) On concrete placement 002-7810-002 at the 310-foot elevation of the Unit 2 Auxiliary Building, embedded foreign material was located with a flex drill.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) interviewed the individuals who made allegations AC-25 and AC-28. A11ega-tions AC-32, AC-33, and AC-34 were made by former Brown & Root employees.
The TRT attempted to determine their identity but was unable to do so. i a. The alleger originally stipulated that the hollow places were located behind the stainless steel liner of Unit I reactor cavity, but when interviewed by the TRT, he stated that he meant Unit 2. In assessing the allegation, the TRT interviewed the TUEC chief structural engi-
- - neer who stated that.when forms were removed from one section of the Unit 2 reactor cavity structure, honeycombed areas were discovered on
' the side of the structure accessible to visual examination. Because of the concern that the honeycombing indicated inadequate concrete consolidation in this section and because the possibility existed K-39 i ,
. . . _ _. .---*e---- + * *
, w , - *- . . . , . - . , . . _ . - . - _ _ , , , - _ .. ., . . ._ _ , -. -.
- d. The existing cracks in the Unit 1 concrete reactor cavity wall have been the subject of a great deal of attention by the NRC and the
, designer. They have been documented in numerous NCRs, such as NCR C-650 and NCR C-1034. The TRT reviewed a random sample of the concrete placement packages for the Unit 1 Containment Building, Auxiliary Building, and both Safeguards Buildings, and found no evidence of specification violations during the concrete placement.
The TRT also inspected the cracks documented in NCRs 1034 and 650. 1 The crack documented by NCR C-1034 is a small hairline crack, caused i by shrinkage or thermal effects, that is so small that it cannot impair structural behavior and capacity. Cracks documented by NCR C-650 are evaluated in Civil and Structural Category 13.
- e. The NRC Resident Reactor Inspector (RRI) at Comanche Peak Steam Electric Station (CPSES) investigated this allegation (Region IV Inspection Report 50-445/80-16 and 50-446/80-16). The RRI learned that the alleger had worked at the site for 5 weeks in early 1980 in the Unit 1 Safeguards Building at the 790-foot elevation. The RRI found two locations at that elevation where a hollow sound could be obtained by tapping a wall with a hammer. He informed TUEC of this condition, and they found several more locations in the same general vicinity, all at the 790-foot elevation. Each area was marked and excavated to approximately 2 inches, that is, to the depth behind the first layer of reinforcing steel. The RRI observed several excavations and saw nothing abnormal about the concrete. He also queried the craft personnel who were excavating when he was not present and was informed that all excavations revealed nothing except uniformly solid concrete. The RRI tapped the concrete after it had been excavated to a depth of approximately 4 inches and could no longer detect a hollow sound. The allegation apparently was based on the premise that what the alleger interpreted as a hollow sound indicated a void in the wall. Excavations of the areas in question revealed no voids in the concrete.
- f. (1) This item was not the subject of an allegation. The TRT reviewed its disposition because it involved an issue similar to those raised in other CPSES allegations.
Nonconformance report (NCR) C-82-00858, which was reviewed by the TRT, indicates that a void did exist in the generator com-partment wall of the Unit 1 Reactor Building. As part of the NCR resolutiori, the matter was reported to Gibbs & Hill (Office Memorandum CPPA-21495, July 20, 1982) and they concluded that the wall would perform both its structural and radiation shield-ing functions whether or not the void was filled. However, to ensure that no safety issue could be raised, Brown & Root filled l the void with nonshrink grout in August 1982, as documented in
- Inspection Report IR-C-6682. The TRT agrees that in its repaired state the wall presents no safety problem.
(2) This item was not the subject of an allegation. The TRT reviewed its disposition because it involved an issue similar to those raised in other CPSES allegations. t K-41 ' l i l
- d. While the allegation of cracking in the concrete basemat is accurate, it is not correct to assume that detrimental structural consequences
. will result from the cracks. The structures are designed to tolerate cracks of the magnitude and location of those found.
- e. The allegation of numerous concrete voids was not substantiated.
- f. (1) The reported void in the generator compartment wall of the Unit 1 Reactor Building is true. The void was filled even though it did not require filling from the standpoint of adequacy of design. The TRT determined that the wall in its repaired ~
condition is safe. (2) The area reported as containing unusual material in the concrete was adequately repaired so that this condition will have no impact on safety. The TRT will inform the individual who made allegation AC-25 of the TRT's findings by letter. The alleger of AC-28 has been notified by letter of the TRT disposition of his allegation. The allegers of AC-32, AC-33 and AC-34 are former B&R employees. The TRT was unable to establish their identity.
- 6. Actions Required: The repairs and the repair documentation to the honeycombing discussed in Item a must be inspected / reviewed and approved by the NRC Resident Inspector before the TRT can determine whether this issue has been adequately resolved. The successful completion of the repairs shall be reported to the TRT and will be verified by the NRC Resident Inspector prior to low power operations.
4 I e K-43
t
- 1. Allegation Category: Civil and Structural 5, Miscellaneous Concrete
- 2. Allegation Number: AC-26, AC-31, AC-36 and AC-43
- 3. Characterization: It is alleged that the following irregularities occurred in connection with concrete construction:
- a. Equipment was set on grout before the grout properly gained strength through aging (AC-26).
- b. Hanger inserts were installed at improper angles (AC-31).
- c. Trash in the bottom of a form was covered with concrete (AC-36). -
AC-43 did not include any new allegations; it merely reiterated those made in AC-26, 31, and 36.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) interviewed the alleger of AC-31. Allegation AC-26 was judged as having sufficient clarity for technical resolution without initial contact between the TRT and the alleger. Allegation AC-36 was the subject of testimony given by Region IV inspectors, but the identity of the alleger was not revealed. The TRT attempted to determine the alleger's identity but could find no record of it.
- a. Allegation AC-26 concerns grouting of stelel plates which were baseplates.for the frames used to support parts of the internal assembly in Unit No. 2 when they were removed from the reactor pressure vessel. If the grout were damaged by the steel plate being loaded prematurely, the damage would occur immediately, while the grout was weak. If the grout survived the loading operation without damage, it probably would not suffer damage in use, since it gains strength rapidly while it is fresh and at a decreasing rate as it ages.
l All elements of the internal assembly were located at the 860-foot. elevation. The TRT inspected all the grouted plates at the 860- and 862-foot elevations and found no evidence of grout failure. While the allegation may be true, all the grout survived the initial loading without damage. If this allegation is true, a quality control issue exists. The TRT Civil and Structural Group did not look into the QA/QC aspects of this allegation.
- b. It is alleged in AC-31 that Richmond anchor bolt inserts were installed between the 860- and 905-foot elevations in Unit 1 at ant'es not perpendicular to the concrete surface and that this condition was compensated for by use of tapered washers. The allegation referred to discrepancies as great as ten degrees. The allegation was addressed in NRC Inspection Report 50-445/83-27, which was reviewed by the TRT.
The TRT found that Brown & Root Procedure CP-CPM 9.10, " Fabrication of ASME-Related Component Supports," stated in Section 3.3.2 that: K-45
- 5. Conclusion and Staff Positions: The TRT concludes that these allegations have no structural safety significance.
- a. All of the grout in question survived the initial load application without failure (AC-26). The possibility of premature loading will be assessed as part of the overall programmatic review concerning procedures addressed under QA/QC Category 6 "QC Inspection."
. b. No infraction of installation procedures for anchor inserts was found (AC-31).
- c. The allegation that trash was dumped into the bottom of a concrete form cannot be substantiated. Even if true, the containment structure concrete, including the dome, was shown to be adequate and acceptable in the in-situ structural integrity test (AC-36).
The TRT scheduled an interview with the alleger of AC-31 to discuss the TRT findings, but he declined to appear. A letter will be sent to him in lieu of a closing interview. The individual who made allegation AC-26 will be informed of the TRT's findings by letter. The TRT could not establish the identity of the individual who made allegation AC-36.
- 6. Actions Required: None.
K-47
- 1. Allegation Category: Civil and Structural 6, Rebar Improperly Installed or Omitted
- 2. Allegation Number: AC-30, AC-37, AQC-12, AC-38, AC-39 and AC-49
- 3. Characterization: It is alleged that reinforcing steel (rebar) was not properly inspected upon receipt at the site (AQC-12 and AC-37). It is also alleged that rebar was omitted in the following locations:
- a. A 6 foot x 6 foot section of concrete in the Safeguards Building
~ ,
(AC-30).
- b. The Unit 1 containment structure wall, specifically horizontal " tie" reinforcement (AC-38).
- c. Four column faces in the wall along column line EA of the Auxiliary Building (AC-39).
- d. It is also alleged that reinforcement was installed upside dowr, in a building near the Unit 2 containment structure (AC-49).
In addition to these allegations, the Region IV resident inspector requested that the NRC Technical Review Team (TRT) review the following possible reportable design deficiencies involving reinforcing steel (rebar):
- e. Reportable Design Deficiency Concerns:
(1) Rebar was omitted in a reactor cavity concrete placement between the 812-foot and 819-foot, 1/2-inch elevations in the Unit 1 Reactor Building. (2) Brown & Root construction requested a change in the configura-tion of two rows by nine layers of No. 9 reinforcing bars (2 x 9
- #9), as shown on drawing 2323-S1-0572, Rev. 4, to a continuous circular arrangement.
(3) Because of interferences with 14-inch diameter sleeves, the horizontal tails of No. 11 vertical reinforcing bars within the triangular columns surrounding the reactor cavity were modified to clear the sleeves. Also because of extreme congestion within the columns, stirrup details were modified. (4) Six No. 10 additional horizontal bars were omitted from a beam above a construction opening on column line KA between 6A and 7A in the Auxiliary Building. (5) Nine No. 9 and two No. 4 additional reinforcing doweis were omitted around an elevator shaft door in the Unit 1 Reactor Building. (6) Forty-six No. 9 dowels on the face of the wall in the excess letdown heat exchanger room in the Unit 1 containment structure were omitted. K-49
- a. During an interview with the alleger, the TRT learned that the allegation of missing rebar in the Safeguards Building actually referred to the return pump station at Squaw Creek Dam (AC-30). For the detailed assessment of this allegation, see Civil and Structural Category 12, AC-29.
- b. This allegation (AC-38) was first reviewed in NRC Region IV Inspection Report No. 79-25, which refers to the omission of horizontal tie rebar in the Unit I containment structure, and concludes that the alleger was referring to an occurrence in the
- Unit 2 contaipment structure rather than in Unit 1. This event occurred shortly before the alleger terminated his employment, and it l, was assumed by the Region IV inspector to be the event to which he referred. The omission of horizontal shear tie reinforcement in Unit 2 was originally investigated in Region IV inspection report 79-18, which notes that this reinforcement had been omitted near the junction of the containment wall and the hemispherical dome and was subsequently placed at a higher elevation. An analysis by Gibbs &
Hill (G&H) concluded that the structure would be capable of carrying
; the design loads with the reinforcement in the as-built location. To determine if the allegation did indeed pertain to the Unit 1 containment structure and if all the reinforcing steel was placed in
, the Unit 1 containment wall as required, the TRT reviewed all 33 concrete pour packages (101-5805-001 through 101-5805-033) pertaining to the main concrete placements in the Unit 1 containment wall. These pour packages contained rebar placement checklists which documented the results of inspections performed by B&R QC confirming the placement of the reinforcing bars to the applicable drawings. The TRT found three placement inspections in which the reinforcing bar placement was initially checked as unsatisfactory; the problems , were then corrected, and the placement was signed off as satisfactory. i The other 30 inspections performed were all checked as being satisfactory in that there were ro deviations from the drawings.
- c. . On October 27, 1977, a nonconformance report (NCR) C-806 was issued reporting the omission of 12 No. 8 vertical wall reinforcing bars at 4 column locations in the wall along column line EA of the Auxiliary Building (AC-39). The reinforcing steel had been omitted between the 810-foot, 6-inch and 831-foot elevations and involved four separate concrete placements made from May to October 1977. This information was submitted to G&H engineering for resolution. G&H performed an
, analysis which showed that the columns remained capable of carrying
- the design loads without the missing reinforcing bars and further
, directed that the bars be omitted from the columns for the remainder of their height through the 873-foot, 6-inch elevation.
- d. The TRT reviewed the April 10, 1979, transcript of a Region IV interview with an alleger and identified an allegation that reinforcement was installed upside down in a building near the Unit 2
- containment structure (AC-49). However, during the interview the alleger claimed that the problem had been corrected prior to concrete placement.
K-51 4
- r. - - - - , - e. , _ - - _ . , ..- ,,,..m,,,,-m .m ,m,,,,_, _ . - . - . , . . , . , . - . - _ _ _
to allow for installation. The stirrup de:;ign was modified to a two piece design rather than one piece, as originclly designed. This modification was permitted only within the triangular columns. (4) On October 26, 1977, nonconformance report (NCR) No. C-809 was issued by Brown & Root reporting the omission of six No. 10 additional horizontal reinforcing bars from a beam over a construction opening on column line KA between 7A and 6A in the Auxilia,ry Building at the 831-foot, 6-inch elevation. G&H engineering issued DC/DDA No. 558 in response to the NCR. G&H engineering stated that the reinforcing bars were not required provided that one of the following conditions was met: (1) shoring remained within the construction opening until the slab above 831-foot, 6-inch elevation and the wall along column line KA above this elevation reached their design strengths, or (2) slab shoring remained adjacent to the construction opening until the concrete used to close the construction opening had reached its design strength. The intent was to provide adequate support to the 831-foot, 6-inch slab from either the wall above, the wall below, or from shoring. The disposition of the NCR showed that the shoring was left in the construction opening until the concrete wall and slab above had cured. The TRT reviewed the design change and solutions proposed and found the approach taken to be satisfactory. The TRT also reviewed drawing SAB-00711, which showed that the construction opening was closed with concrete pour No. 002-4810-042 on January 30, 1979. (5) Brown & Root issued NCR No. C-810 reporting the omission of nine No. 9 and two No. 4 additional reinforcing dowels around the elevator shaft door in the Unit 1 Reactor Building at the 832-foot, 6-inch elevation. G&H DC/DDA No. 477 indicated that the nine No. 9 dowels were to be drilled and grouted in place, and that the two No. 4 dowels could be placed without doweling into the slab. A review of the safety implications of the omitted reinforcing bars by Texas Utilities Electric Company (TVEC) Design Engineering showed that cracking of the concrete in this area could have occurred during conditions such as a seismic event if the reinforcing steel had not been placed. The review concluded that the cracking would not have affected the safety of the structure. (6) On October 31, 1977, NCR C-811 was issued by Brown & Root reporting the omission of 46 No. 9 dowels on the face of the wall in the Excess Letdown Heat Exchanger Room in the Unit 1 Reactor Building. The civil QC inspector involved stated that the reinforcing steel had been installed and checked but that it was subsequently removed to allow for the installation of steam generator lower supports and reactor coolant pump tie supports and not replaced. G&H engineering directed that the dowels be drilled and grouted in place. K-53
1 i reinforcing steel installation as per DC/DDA 518 Rev. 1 was i inspected and accepted and that the concrete was placed on November 11, 1977. Therefore, the TRT concluded that the correct bars were used. The six documented structural sections with omitted reinforcing steel above indicate a breakdown in the quality contrcl program as evidenced by the fact that these omissions were not detected prior to concrete placement.
- 5. Conclusion and Staff Positions: For the allegat'ons concerning improperly inspected rebar (AQC-12, AC-37), the TR1 concludes, based on the fact that
- the reinforcir.g steel used was subsequer.t'.y accepted by QC, that this issue has no effect on the structural safety of the structure.
The TRT reached the following conclusions for the remaining allegations ] and reportable design deficiencies:
- a. Allegation AC-30, which refers to the return pump station at Squaw Creek Dam, not the Safeguards BuildiLg, is examined in Civil and Structural Category 12, AC-29.
- b. For AC-38, the TRT concludes that the horizontal shear bar reinforcement was placed in the Unit 1 Containment Building wall as required and further agrees with the conclusion drawn in Region IV Inspection Report No. 79-25 that the allegation refers to the Unit 2 containment structure, where the G&H analysis showed that the
. structure would be capable of carrying the design loading with the
, reinforcing steel in its as-built location. Therefore, the TRT concludes that this issue has no structural safety significance.
- c. The TRT reviewed the G&H analysis and agrees with their methodology and conclusion (AC-39). The TRT, therefore, concludes that this l allegation has no structural safety significance.
- d. The TRT concludes that since this instance of improperly installed
! rebar was corrected prior to concrete placement, this issue has no adverse effect on the struct ral safety of the structure.
- e. (1) TheTRTcannotdeterminkthesafetysignificanceofthisissue i until an analysis is performed verifying that the reinforcing steel in the as-built condition is adequate.
(2) The TRT concludes that the change made to the No. 9 reinforcing i bars did not affect the load-carrying capacity of the structure. (3) The TRT finds the modifications made to the interfering bars to i be acceptable and to have no adverse effect on the structural safety of the structures. l (4) The TRT finds that the omission of the additional reinforcing
- bars will have no adverse effect on the structural safety of the structure because shoring left in place until the concrete had cured made the additional reinforcing steel unnecessary.
K-55
-g_ .p.
_ _ . . . - . - . _ _ _ _ m __ _ . _ - _ _ . - _ ._.
l I
- 1. Allegation Category: Civil and Structural 7, Uncontrolled Repair
- 2. Allegation Number: AC-10
- 3. Characterization: It is alleged that the removal of a Hilti bolt from the floor at the 852-foot level of the Safeguards Building resulted in a cone shaped section of concrete being removed which was later repaired in an " uncontrolled manner."
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT dI'd not initially attempt to contact the alleger because the allegatiod) was sufficiently clear for the TRT to proceed with its investigation.
In assessing this allegation, the TRT examined NRC Investigation Report 81-12 (April 16, 1982), which described the observations of the area in question by an NRC investigator and the senior resident inspector. They concluded that the floor was repaired with a surface patch rather than being repaired all the way through. Such an uncontrolled and undocumente repair of a portion of a Category I structure indicates a lack of QA/QC control. The TRT concurred with these findings based on its observations of the floor area in question. Nevertheless, the TRT performed an independent evaluation of the safety significance of a 14-inch-diameter hole extending through the floor slab adjacent to pipe support No. CC-1-137-700-E63R, as alleged. This hole is located in the Electrical and Control Building and not in the Safeguards Building, as alleged, and as reported in NRC Investigation Report 81-12. For the worst-case analysis, the TRT assumed that two reinforcing steel bars (rebars) were cut in the process of removing the Hilti bolt. To account for the unknown quality of the material used in the repair, the TRT computed the ultimate moment capacity of the floor slab with and without a 14-inch section of slab removed. These estimated strength capacities were compared to the strength requirements necessary to resist the actual moments resulting from the slab design loads. The adequacy of shear capacity was also verified in a similar manner. From these analyses, it was evident to the TRT that the slab in its as-built condi-tion is capable of resisting the actual design loads, even though the most conservative engineering assumptions concerning cut rebar and a 14-inch hole were made.
- 5. Conclusion and Staff Position: Based on observations made by the TRT, the floor slab does not show any sign of degraded capacity or of p.cor repair practices. The slab appears continuous and composed of good materials.
An independent TRT analysis of the slab capacity, based on conservative engineering assumptions, confirmed that the structural integrity of the slab would be maintained under its design loads. Accordingly, this alle-gation.has no structural safety significance. l K-57
?
- 1. Allegation Category: Civil and Structural 8, False / Wrong Documents
- 2. Allegation Number: AQC-1, AQC-2, AQC-3, AQC-7, AQC-46 and AQC-51 -
- 3. Characterization: It is alleged that the following records were falsified at various times:
- a. Concrete air entrainment records (AQC-1).
- b. Concrete laboratory test records (AQC-2). This allegation consisted of four separate parts: (1) that slump records were falsified, (2) that laboratory tests (air, slump, and temperature) for concrete placements of 10 cubic yards or less, prior to 1978, were not performed, (3) that laboratory tests were signed by a Level II inspector not present at the time the tests were performed, and (4) that the alleger signed a pressure gauge qualification test that he was not qualified to certify.
- c. Aggregate tests (January 1976). The alleger maintains that he and
- his foreman falsified these tests (AQC-3).
- d. Compression strength tests, at the direction of the general foreman ,
and laboratory manager (AQC-7).
- e. Midpour tests during the placement of the Unit 1 Containment Building basemat on February 21, 1976 (AQC-46).
i
- f. Cadweld tensile test records were reported by an inspector without the tests actually being performed during the spring and summer of 1976 (AQC-51).
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) attempted to determine the identities of the individuals who made allega-tions AQC-1 and AQC-2 and could not find any record of their identities.
Allegation AQC-46 appeared in a April 1979. Fort Worth Star-Telegram article; the allegation was made by three unidentified individuals. The TRT did not initially attempt to contact the individual who made allega- - tions AQC-3, AQC-7, and AQC-51 because the allegations were sufficiently : clear to allow the TRT to proceed with its investigation.
- a. In assessing this allegation (AQC-1), the TRT reviewed documents ,
contained in Brown & Root (B&R) Deficiency and Disposition Report (DDR) No C-488 R1, R. W. Hunt Company QA Report HCP 21697 on concrete acceptance test results and the results of the Region IV investiga- l tion of this allegation (inspection report 77-02). The records showed that on January 20, 1977, a 3.9 percent air content value was : recorded in the concrete acceptance test report (HCP 21697) as 4.3 percent by a Level I inspector. The incident was reported to R.W.
- Hunt management by a co-worker. R. W. Hunt then issued a DDR ,
identifying the placement of out-of-specification concrete and " corrected the air entrainment value in the acceptance test report. The Level I inspector was subsequently fired for his action. J 1 i i K-59 i
-n .- - - - - - - . . - ,, , - . -.
,,.,n n, . .
n - - - - - - - - - - . - ,.----- - - - -
.- - - - = . - .
J another employer. Three of them were working in the concrete testing laboratory. All four stated that they did not participate in or observe any falsification and/or failure to perform required concrete tests. (3) The allegation that a Level II inspector signed reports for tests performed on September 3 and 4,1977, that he could have had no knowledge of was also reviewed by NRC Region IV personnel (IE Inspection Report 78-07). The alleger stated he had obtained this.information from another individual who thought the falsification occurred in December 1977. The Region IV inspectors reviewed the daily payroll records of all laboratory personnel for the first 10 days of September and all of December 1977. The Level II inspector was present every day in September, but was absent December 4, 5, 11 and 18 through 31. The Region IV staff could find no reports validated by the Level II inspector for the days alleged. The TRT reviewed the Region ' IV inspection report and concurred with the approach taken and the results of the investigation. In addition to reviewing the Region IV inspection report, the TRT examined strength test results for the concrete placed during the period stated.in the j allegation and found them to be above the minimum required design strength. (4) The allegation that the alleger signed a pressure certification ! test that he was not qualified to certify (on August 15, 1977) was investigated by NRC Region IV personnel (IE Inspection Report 78-07). Through an interview with Brown & Root calibration facility personnel, the NRC Region IV investigator i learned that the pressure gauge was calibrated by Brown & Root personnel .1 accordance with their procedures. The calibration record was an R. W. Hunt form signed by the alleger who observed the test in accordance with the R. W. Hunt procedure. Prior to the Region IV investigation, B&R issued a DDR (February 17, 1978) that described this situation as a pre-existing and continuing problem in general, and proposed corrective action. l The NRC Region IV staff concluded that while the allegation was substantiated, there were no safety consequences since the calibration was performed by a qualified individual in accordance with prescribed procedures. The TRT reviewed the Region IV inspection report, and concurred with the approach taken and the results of the investigation. -
- c. The allegation (AQC-3) was first evaluated by the NRC Region IV staff (IE Inspection Report 79-09). The alleger, a former R. W. Hunt 1 employee, stated that the falsification by him and his " foreman" occurred during the first 3 or 4 weeks of his employment, beginning January 19, 1976. The NRC Region IV staff reviewed the pre-qualification tests performed by Texas Industries, the aggregate supplier, on the material supplied to the site between January and l May 1976 and also examined the results of in process concrete testing. Both sets of results complied with the specification requirements. The NRC Region IV staff also determined through discussions with a TUEC representative that the " foreman" was a Level K-61 O K
- 4.-. .y . . , _ . , , ,.. , , _ , . . , . _ . ..
1 1 I concrete cylinders would be the final measure of its acceptability, the TRT reviewed these results and found them to be acceptable and within the specification.
- f. The TRT reviewed all 440 Cadweld tensile test results for 1976 and identified 30 tests that were performed by the inspector in question (AQC-51). Twenty-eight of these tests were performed on one single day (October 13, 1976), while the other two tests were performed on two different days (July 21, 1976 and August 20, 1976). The TRT cannot determine whether or not all of the 30 tests in question were p'erformed or if results were falsified and did not specifically look into the falsification issue. The remaining 410 tests performed by other inspectors all met the tensile strength requirements. The 30 Cadwelds tested were removed from the first layer of the exterior wall of the Unit 1 Containment Building at the 832-foot, 6-inch elevation. The TRT reviewed tensile test results of other Cadwelds performed by the individuals who made the Cadwelds in question. The results were found to be satisfactory. The Cadweld rejection rate for the 21 Cadwelders who made the 440 Cadwelds ranged from zero percent to four percent, with one at six percent.
The fact that the allegations concerning the falsification of an air entrainment test and the certification of the pressure gauge test were substantiated indicates a partial breakdown in the QA/QC program in these areas.
- 5. Conclusions and Staff Position: The allegation (AQC-1) that a concrete air entrainment record was falsified is true. Even so, the compressive strength of the concrete in question was within specifications.
The allegation (AQC-2) that slump tests on April 11 and 13, 1978, were performed incorrectly and that the results were falsified could well be true and cannot be refuted. The TRT examined the compressive strength test results of the concrete in question and found that they were within specifications. The allegation (AQC-2) that laboratory tests for small placements were falsified was found to have no structural safety significance since, in j addition to the recorded laboratory tests, the validity of which was l questioned, cylinder strength tests were also performed to demonstrate l adequate strength. In addition, in interviews with the TRT, former l employees of the R. W. Hunt Co., who worked during the time period cited in the allegations, denied the validity of the allegation. Furthermore, the limited number of concrete placements of less than 10 cubic yards, l even if improperly tested, would have little structural safety l significance. The allegation (AQC-2) that an inspector signed test results for which he could have had no knowledge could not be substantiated because no reports ! could be found which had been signed by the inspector on the days alleged. l Even if the allegation were true, test results showed the strength of the
. concrete placed during the period of the allegation to be above the minimum required strength.
K-63
- 6. Actions Required: TUEC shall determine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a progran to assure acceptable concrete strength. The program shall include tests such as the use of random Schmidt hammer tests on the concrete in areas where safety is critical. The program shall include a comparison of the results with the results of tests performed on concrete of the same design strength in areas where the strength of the concrete is not questioned, to determine if any significant variance in strength occurs.
TUEC shall submit the program for performing these tests to the NRC for review and approval prior to performing the tests. e 4 K-65
- 1. Allegation Category: Civil and Structural 9, QC Inspector Training
- 2. Allegation Number: AQC-9
- 3. Characterization: It is alleged by two former R. W. Hunt Company employees that (a) after a March 1977, NRC investigation, closed book recertification tests of R. W. Hunt inspectors were done "open book" and that (b) tests were given with the answers provided.
- 4. Assessment of Safety Significance:
The NRC Technical Review Team (TRT) did not initially attempt to contact the two allegers because the TRT was able, with some initial investigation, to clarify the allegations suffi-ciently to proceed with its evaluation. The allegation about the recertification tests refers to recertification testing that was required because a Region IV investigation in 1977 (inspection report 50-445/77-02) questioned the most recent certification of R. W. Hunt Level I and II inspectors. The NRC Region IV staff found that R. W. Hunt did not comply with the minimum 2 year experience requirement for qualification as a Level I concrete inspector, as required by the ASME Code to which they were committed by the Preliminary Safety Analysis Report (PSAR). The R. W. Hunt Skills Training Certification manual stated that " Experience requirements may be reduced if the individual can demonstrate capability in a given job through previous performance or satisfactory completion of an examination and orientation training." Alsc, the Region IV staff found that the " Certification of Qualifications," which was issued to each Level I and II inspector did not include the activity the inspector was qualified to perform or the basis used for certification, as was required. Each candidate for certification was required to demonstrate proficiency in performing specific practical tests on one or more samples approved by the Level III examiner. The Region IV staff found that R. W. Hunt had permitted a Level I inspector to perform concrete cylinder compression tests and aggregate sieve analysis without evidence of demonstrated proficiency and approval in accordance with the above requirements. As a result of this investigation, Brown & Root (B&R) audited R. W. Hunt training and certification activities and required each inspector to be recertified by attending specific training sessions and by closed book testing. The work performed by the personnel qualified under the previous provisions was reviewed by B&R QA personnel and found to be within the specification requirements. In addition B&R assigned a QC civil engineer to work full-time with R. W. Hunt on site to ensure full compliance with project requirements. Between April 3, 1979, and May 7, 1979, NRC Region IV inspectors inter-viewed 15 individuals associated with concrete testing activities regard-ing the allegation concerning the recertification tests (inspection report 50-445/79-09), including the 2 who had originally made the newspaper l allegations (April 1979). The alleger who stated in the newspaper article that after the NRC investigation of March 1977, the recertification of K-67 l
certified prior to March 1977 was reviewed by B&R and was found to be within specifications, a fact subsequently reported to NRC Region IV staff. Therefore, the TRT concludes that this allegation has no structural safety significance. The allegation that the recertification tests for concrete cylinder testing were given "open book" cannot be substantiated. This allegation was not supported by any of the other individuals questioned, which sug-gests it was an isolated occurrence. The work performed by this individual prior to March 1977 was audited and found to be satisfactory, which would indicate the individual possessed the knowledge required to properly per-form the required testing. In addition, the test results for concrete placed, including the concrete compression tests, were contributed to by many inspectors whose qualifications were acceptable. These test results showed the concrete was of uniform quality and strength. Based on the fact that the inspector's work had previously been reviewed and found to be acceptable, and that a number of inspectors contributed to the test results, which showed the concrete to be of uniform quality, the TRT concludes that this issue has no structural safety significance. The results of these evaluations will be further assessed as a part of the overall programmatic review concerning inspector qualifications addressed under QA/QC Category 4, " Training and Qualification of Personnel." There-fore, the final acceptability of these evaluations will be predicated on the satisfactory results of the programmatic review of this subject. adjustments'to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER. The TRT contacted one of the allegers who made one of the allegations discussed above. He declined to meet with the TRT. He will be informed by letter of the TRT's findings. The TRT is attempting to contact the other alleger involved.
- 6. Actions Required: None.
l K-69 i l 1
1 l
- 1. Allegation Category: Civil and Structural 10, Improper Testing
- 2. Allegation Number: AQC-4, AQC-5, AQC-6, AQC-8, AQC-11 and AQC-48
- 3. Characterization: It is alleged that the following violation of testing procedures occurred:
- a. Equipment required for aggregate testing was sitting unused on laboratory shelves (AQC-4).
- b. Shortcuts were taken on tests involving grading of aggr,egate (AQC-5). ,
- c. During the placing of' a 6600-cubic yard section of the basemat for Unit 1, some concrete was placed without the required testing (AQC-6).
- d. Concrete cylinder compression tests were run at a faster loading rate than permitted by NRC regulations (AQC-8).
- e. Concrete test cylinders with adequate strength were used to represent other placements (AQC-11).
- f. Concrete test cylinders in the Hunt Laboratory moist room were allowed to dry (AQC-48).
Allegations AQC-4, AQC-5, AQC-6, and AQC-8 were investigated by NRC Region IV and documented in inspection report 79-09, which was reviewed by the NRC Technical Review Team (TRT) as a step in its own assessment of the allegations.
- 4. Assessment of Safety Significance: Allegations AQC-4, AQC-11, and AQC-48 were judged as having sufficient clarity for technical resolution without
- initial contact between the TRT and the alleger. Allegations AQC-5, AQC-6, and AQC-8 were made in newspaper articles which did not identify the allegers, and the TRT has been unable to determine the allegers' identities.
- a. The test equipment that allegedly remained unused at the project laboratory was for the test for Potential Reactivity of Aggregates (Chemical Method), American Society for Testing and Materials (ASTM)
C 289. Test Laboratcry Manual TLM-004 (CP-QP-0.5), which was in ! effect during most of the construction period, required that the test l be run once for each 4000 tons of aggregate. The TRT inspected l " Folder 1 - Potential Reactivity, 4000 Ton Test" and learned that I between May 6, 1975, and July 12, 1978, there were 60 tests for potential reactivity. The TRT also interviewed the laboratory tech-nician who performed most of the tests. This period covers the bulk of heavy construction and the entire employment period of the alleger. The testing rate during this period exceeded one test per 4000 tons of coarse aggregate. Thus, testing was at a higher rate than required by the testing requirements. t . K-71
completed, at most, two extra cylinders remained for which the test , results could be switched to the testing data for other placements. I Unless this was a widespread practice, its significance would be I small because of the relatively few cylinders available. The I allegation does, however, raise a question as to the effectiveness of l quality control in the laboratory. I
- f. To investigate the allegation that concrete cylinders in the laboratory moist room were allowed to dry, the TRT exam.ned the !
current procedure for documenting moist room conditions and interviewed a Level II inspector who was present throughout the period when the laboratory was operated by the R. W. Hunt Company. 1 At present there is a thermometer which provides a permanent I temperature record. While there is no quantitative measurement of humidity, there are daily visual observations of the presence or absence of fog in the room. These observations are also a part of the record. During the R. W. Hunt operation, temperatures were recorded, but there apparently was no documentation of humidity. Batch plant inspectors were required to note the condition of the moist room, but there is no record of their observations. There is a history of breakdowns in the water supply to the laboratory, and a shutdown as long as 6 hours has been documented. With the door to the moist room closed, there would be a negligible drop in humidity during such a period. As long as the relative humidity remains above 90 percent, concrete curing conditions are favorable. It is pertinent to note 4 that any drying that might occur would produce conservative results in that measured strengths would be lower than actual strengths.
- 5. Conclusion and Staff Positions: The TRT concludes that these allegations have no impact on structural safety.
- a. All required tests for ASTM C-289 were performed.
- b. The alleged shortcut in carrying out aggregate grading tests is permitted by the provisions of the specified test method in ASTM C-136.
- c. All required testing was carried out in connection with the 6600-cubic yard basemat placement.
- d. Although this allegation may have been true, the fastest possible loading of test cylinders would have increased the indicated strengths by no more than 6.5 percent and would have had no effect on the ccceptability of the concrete.
- e. The alleged substitution of test cylinders is unlikely to have affected a sufficient number of cylinders to have had a material effect on the overall test results. .
- f. Although the allegation that the laboratory failed to maintain the
. water supply at all times may be true in that there were brief K-73
1
- 1. Allegation Category: Civil and Structural 11, Seismic Design / Construction
- 2. Allegation Number: AC-41
- 3. Characterization: It is alleged that there was poor workmanship regarding the use of elastic joint filler material ("rotofoam") as a temporary spacer during construction to maintain the required air space between
. seismic Category I structures.
- 4. Assessment of Safety Significance: This allegation was received anony-mously; therefore, the TRT could not contact the alleger about its evalua-tion of AC-41. ,
4 TUEC informed NRC Region IV on November 23, 1977, of this allegation, l' which TUEC received anonymously in a telephone call on November 22, 1977. A Region IV inspector reviewed the allegation during an inspection con- , ducted between November 28 and December 2, 1977, and concluded, based on the information available to him at the time, that all temporary rotofoam had been removed from the seismic gap between Category I structures. The matter was left open pending a Region IV review of the Brown & Root (B&R)
- QA/QC inspection and documentation program, which was being initiated to 1 assure that the required seismic gap between Category I structures was being maintained. Rotofoam was used as a temporary spacer during cons-truction to maintain this gap. Once the concrete hardened, the rotofoam t
was removed to eliminate any load transfer or dynamic interaction between buildings. If the relative motion bet' ween buildings was small and the presence of rotofoam was considered in the dynamic analysis of the build-ing, leaving the rotofoam in place may not have had a significant impact l on the dynamic performance of the buildings. During an inspection between January 3 and 13, 1978, the Region IV inspector reviewed B&R procedure CP-QCI-2.4-9, " Inspection of Elastic Joint Filler Material Removal," Revision 1 (December 12, 1977), and B&R inspection reports for December 15, 1977, and January 3, 1978, and had no further questions regarding this matter. l The NRC Technical Review Team (TRT) attempted to obtain a further ; 4 clarification of the concerns expressed by the alleger; however, neither 8 l TUEC nor the Region IV office had records of the alleger's telephone f
- conversation other than what is stated above. The TRT determined,
- however, that prior to the time the allegation was made there was a
, misunderstanding as to whether or not the rotofoam should remain in place l as part of the final construction. A letter from Gibbs & Hill (G&H) of September 6, 1977 (GTT-1543), indicated that construction was improperly . proceeding on the basis that the rotofoam could be left in place. The letter further stated that this assumption was not in accordance with i the facility design drawings and design concept and that expansion joints above grade should consist of a clear gap between buildings, i.e., free of ,
rotofoam. As noted in the G&H letter, it was intended that the rotofoam t i be left in place below grade. Since construction had proceeded above grade, TUEC instructed B&R, in a letter of October 7,1977 (TUS-5012), to ,
- remove the rotofoam above grade. As noted, B&R procedure CP-QCI-2.4-9 was
)
also implemented to verify removal of the rotofoam. Based on discussions K-75
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and the Auxiliary Building at the 830-foot, 6-inch elevation, the air gap was clear to an observer looking up. However, a wooden board and other debris were observed when viewed straight in and downward.
- 5. Conclusion and Staff Positions: Based on the review of available inspection reports and related docLments, on field observations, and on discussions with TUEC engineers, the TRT cannot determine whether an adequate air gap has been provided between concrete structures. Field investigations by B&R QC inspectors indicated unsatisfactory conditions due to the presence of debris in the air gap, such as wood wedges, rocks, clumps of concrete and rotofoam. The disposition of the NCR relating to this matter states that the " field investigation r.eveals that most of the material has been removed." However, the TRT cannot determine from this report (NCR C-83-01067) the extent and location of the debris remaining between the structures.
Based on discussions with TUEC engineers, it is the TRT's understanding that field investigations were made but that no permanent records were maintained. In addition, it is not apparent that the permanent installation of elastic joint filler material ("rotofoam") between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the seismic analysis assumptions and dynamic models used to analyze the buildings, as these analyses are delineated in the Final Safety Analysis Report (FSAR). The TRT, therefore, concludes that TUEC has not adequately demonstrated compliance with FSAR Sections 3.8.1.1.1, 3.8.4.5.1, and 3.7.B.2.8, which require separation of seismic Category I buildings to prevent seismic interaction during an earthquake. Depending on the extent of nonconformance with FSAR Sections 3.8.1.1.1 3.8.4.5.1, and 3.7.B.2.8, the allegation is judged to have merit and potential safety significance. Prompt remedial actiono as delineated below should be implemented. No closing interview could be held regarding this allegation, because the allegation was received anonymously. -
- 6. Actions Required: TUEC shall: -
- 1. Perform an inspection of the as-built condition to confirm that
- adequate separation for all seismic Category I structures has been l provided.
- 2. Provide the results of analyses which demonstrate that the presence
- of rotofoam and other debris between all concrete structures (as determined by inspections of the as-built conditions) does not result i in any significant increase in seismic response or alter the dynamic
! response characteristics of the Category I structures, components, and ( piping when compared with the results of the original analyses. l t K-77 l
, - , er - - -
1
- 1. Allegation Category: Civil and Structural 12, Concrete Construction and 1:eficiencies/ Tolerances
- 2. Allegation Number: AC-29
- 3. Characterization: It is alleged that a spillway pillar, span, or column-was erected 75 degrees to 80 degrees offset and that reinforcing steel t was omitted from a concrete wall.
4.' Assessment of Safety Significance: There are two spillways at Comanche Peak Steam Electric Station (CPSES). One, the service water discharge spillway, is located near the safe shutdown impoundment (SSI); the other is located at the Squaw Creek Dam. The alleger stated that the construc-tion in question took place some time between 1976 and 1977. The spillway ' at Squaw Creek Dam was constructed between August 1976 and January 1977, so it was considered to be the spillway in question. The spillway at Squaw Creek Dam, however, does not have a span, column or pillar. There-fore, on August 3, 1984, the NRC Technical Review Team (TRT) interviewed the alleger to clarify this allegation. From the interview, the TRT learned that the spillway pillar, span, or column to which the alleger referred was located in the Service Outlet 4 Structure below the Squaw Creek Das Spillway, which does have a suspended structure and supports that could be described as a span and pillars. , The TRT inspected the service outlet structure at the Squaw Creek Dam Spillway and found no evidence of any spillway pillar, span, or column which was erected 75 to 80 degrees offset. The TRT also determined that the general configuration of the structure was consistent with that shown on the following drawings: FN-SCR-37 FN-SCR-48
- FN-SCR-39 FN-SCR-49 i
FN-SCR-40 FN-SCR-71 FN-SCR-42 FN-SCR-72 i FN-SCR-44 The TRT learned during an interview with the alleger that the allegation concerning the 6-foot by 6-foot concrete wall area of the Safeguards Building, which allegedly had no reinforcement placed around a pipe j approximately 24-inches wide, was incorrect. (Refer to Civil and Structural Category 6, Allegation AC-30.) The a11eger identified the 6-foot by 6-foot concrete wall area as located in a structure near the Squaw Creek Dam Spillway. The TRT inspected the structures located near the Squaw Creek Dam I spillway and found two structures with a 2- to 3-foot-diameter pipe I surrounded by reinforced concrete. One of these was the outlet works conduit section; the other was the return pump station. The TRT examined 141 concrete placement cards associated with these two structures. K-79
,n_.,--- . , - - - - - - . - - . , - - - . _ _ - - , . . .
- 1. Allegation Category: Civil and Structural 13, Cracks in Concrete Beneath the Reactor Vessel
- 2. Allegation Number: AC-44
- 3. Characterization: It is alleged that detrimental cracks exist in the concrete pad at the bottom of the reactor vessel.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) did not initially attempt to contact the alleger because the allegation
, was sufficiently clear to the TRT to proceed with the investigation.
The existence of these cracks is documented in Nonconformance Report (NCR) C-650. The cracks are in a lift of concrete near the bottom of the reactor vessel, but not in the basemat. The TRT examined concrete placement pack-age 101-2812-001 and NCR C-650, and found no documented violations of specifications during the concrete placement, which occurred on March 21, 1977. Subsequent to placement, however, vertical cracks occurred that extended horizontally to the edge of the reactor cavity. The Gibbs & Hill (G&H) design engineer stated on May 11, 1977, that he found the cracks during his investigation. He attributed the cracks to the mass, configura-tion, and formwork on the interior circumferential face, all of which precluded normal shrinkage, and stated that the cracks were of no struc-tural significance. An NRC Region IV structural engineer also presented his evaluation of the cracks during the ASLB hearing conducted on June 9, 1982. The TRT reviewed this testimony, along with that of G&H engineers given on June 7 and 2nd 8, 1982, and agreed with the assessment contained in their testimony. The doughnut shape of the concrete section and the rigid form in the opening made it virtually impossible to avoid cracking if the entire section was placed in one pour, as it was. However, the structure was adequately reinforced so that the cracks would not impair structural behavior and capacity. The cracks have been repaired at the surface with epoxy resin for operational, rather than structural, reasons. The TRT inspected the concrete and found it to be in excellent condition. The TRT review of the design indicated that the concrete section was originallpdesignedastwosections,withconstructionjointsatthe locations; where the cracks occurred. The contractor was given the option
- of placing the concrete either in two sections with construction joints,
! or in one section without joints. The cracks that formed were not greatly I different from^the construction joint which would have been present if the two placement option had been adopted; thus, the concrete in place i essentially conformed with the original design. One of the cracks was near the mid-span of a deep beam spanning a 20-foot cavity. Reinforced concrete beams must crack in the bottom tensile zone when load is applied. If flexural stresses were kept below the tensile strength of concrete, less than 20 percent of the strength of the steel would be utilized. In design, the reinforcement is distributed so that , the cracks are numerous and very narrow, both for the sake of appearance l and to prevent corrosion of the steel. The occurrence of a pre-existing l K-81
. a
- 1. Allegation Category: Civil and Structural 14, Control Room Area Deficiencies
?
- 2. Allegation Number: AE-17
- 3. Characterization: It is alleged that the field run conduit, the drywall, and the lighting installed in the area above the ceiling panels in the control room are classified as non-seismic and are supported only by wires and that these items may fall as a result of a seismic event.
- 4. Assessment of Allegation: The NRC Technical Rev(ew Team (TRT) did not
- initially attempt to contact the alleger because the allegation was sufficiently clear to allow the TRT to proceed with its investigation.
The TRT electrical group reviewed the electrical aspects of this allegation. (See Electrical and Instrumentation Category 4.) The Civil and Mechanical group of the TRT evaluated the seismic aspects of this allegation. General Design Criteria No. 19 requires that safe occupancy of the control room during abnormal conditions be provided for in its design. The Comanche Peak Steam Electric Station (CPSES) control room is in a seismic Category I structure, with certain seismic Category II and nonseismic components located in the. ceiling. Seismic Category I refers to those systems or components which must remain functional in the event of an earthquake. Seismic Category II refers to those systems or components whose continued functioning is not required, but whose failure could reduce the functioning of any seismic Category I system or component (as defined in Regulatory Guide 1.29) to an unacceptable level or could result in an incapacitating injury to occupants of the control room. Seismic Category II systems or components are, therefore, designed and constructed so that a Safe Shutdown Earthquake (SSE) will not cause such failure or injury. In assessing this allegation, the TRT reviewed the CPSES nonsafety-related conduit, lighting fixtures, and the suspended ceilings installed in the control room. Three types of suspended ceiling exist in the control room: i drywall, louvered, and acoustical. The following list designates those ceiling elements present in the control room and their seismic category designation: j 1. Heating, Ventilating and Air Conditioning - Seismic Category I i 2. Safety-related Conduits - Seismic Category I l 3. Nonsafety-related Conduits - Seismic Category II l 4. Lighting Fixtures - Seismic Category II l 5. Sloping Suspended Drywall Ceiling - Nonseismic
- 6. Acoustical Suspended Ceiling - Nonseismic
- 7. Louvered Suspended ceiling - Nonseismic The TRT also examined the control room ceiling system and pertinent design drawings and met with cognizant Texas Utilities Electric Company (TVEC) engineers on July 31, 1984, to discuss the specific seismic analyses per-i formed for the ceiling elements. In addition, the TRT held a conference call on August 1, 1984, with principal Gibbs & Hill (G&H) design engineers i
l l K-83 l
~
l -
i The TRT concludes that calculations supporting the seismic Category II lighting fixtures do not adequately reflect the rotational interaction with the nonseismic items. In addition, the fundamental frequencies of the supported masses were not calculated to determine the influence of the seismic response spectrum at the control room ceiling elevation. Theindividualwhomadetheallegationdiscussedabohewillbecontacted by the TRT upon resolution of this issue to inform him of the action taken.
- 6. Actions Required: TUEC shall provide:
- 1. The results of seismic analysis which demonstrate that the nonseismic items in the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. An evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems.
- 3. Verification that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
- 4. The results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
- 5. The results of an analysis which demonstrate that the foregoing problems are not applicable to other Category II and nonseismic structures, systems, and components elsewhere in the plant.
l l K-85
- 1. Allegation Category: Civil and Structural 15, Rebar Improperly Drilled
- 2. Allegation Number: AC-13, AC-14, AC-15, AC-18 and AC-40
- 3. Characterization: It is alleged that undocumented and unauthorized holes were drilled through reinforcing steel (rebar). The issue includes allegations relating to:
- a. the loan of rebar drills without proper documentation (AC-13),
- b. the unauthorized cutting of rebar in non-specific locations (AC-14, AC-18, AC-40), and
- c. the unauthorized cutting of rebar used in the installation of the trolley process aisle rails in the Fuel Handling Building (AC-15).
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) contacted the individual who made allegations AC-13 and AC-14 to clarify his concerns. The TRT did not initially attempt to contact the individuals who made allegations AC-18, AC-40, and AC-15.
- a. AC-13 concerns the loan of rebar drills allegedly used for the unauthorized cutting of rebar. During the NRC investigation of this _
matter, the NRC Office of Investigation (OI) interviewed nine j individuals alleged to have knowledge of unauthorized cutting of y rebar. These individuals provided sworn statements denying any i knowledge of this activity. These statements are a part of OI Report A4-83-005 (May 20, 1983), which concludes that "there was no testimony received indicating that holes were drilled or rebar was cut without proper documentation, and no evidence was found to contradict the testimony of these individuals." One instance of possible unauthorized cutting of rebar is discussed in a supplement to the OI report (September 7, 1983). This instance is discussed below in relation to allegation AC-15. Because the alleger did not specifically identify who made unauthorized cp4 df rebar, or where this cutting took place, the TRT attempted to Oant' 'y the amount of rebar that allegedly was cut without aith / v '2n. In discussions with the TRT, the alleger estimated Lt..it wximately five percent of the diamond core drill bits ordered by him were used in an unauthorized manner. He further estimated that one drill could be used to cut up to five rebars, depending upon the extent of cutting required. Although he could not be specific as to how many drills he ordered, the alleger thought that the number would be in the thousands. The NRC Region IV Investigation of this issue indicated that 415 diamond core drill bits were purchased during the period in question (IE Report 83-27). Using the actual number of drill bits purchased, together with the information provided by the alleger, the TRT estimated that there could be approximately 100 alleged unauthorized rebar cuts. Considering the large amount of reinforcing steel used in the plant, and the fact that the structures consist primarily of heavily reinforced concrete walls and slabs, the TRT determined that, if such K-87 ,
- c. Allegation AC-15 identifies a specific instance of the possible unauthorized cutting of rebar. In this case, a former Brown & Root employee stated he possibly drilled holes through rebar in a concrete floor without a component modification card (CMC) or a design change authorization (DCA). He explained that in January 1983 he drilled approximately 10 holes about 9 inches deep while installing 22 metal plates with a core drill. He said the metal plates were used to secure the trolley process aisle rails located on the S10-foot, 6-inch floor level in Room 252 of the Fuel Handling Buildir.g.
The TRT inspected the trolley process aisle rails and its anchoring system and observed no violations of project drawings or specifications. The TRT reviewed the reinforcement drawings (2323-S-0800 and 2323-S-0820) for the Fuel Handling Building to determine the location of rebar. The drawing showed three layers of reinforcement in the upper part of the mat, which consisted of a No. 18 bar running in the east-west direction, in the first and third layers, and a No. 11 bar running in the north-south direction, in the second layer. The review of the reinforcement drawings (2323-S-0800 and 2323-S-0820) revealed that the layout of the east-west reinforcement and the
- trolley process aisle rails was such that only one bar of the east-west reinforcement could be cut by drilling holes for rail anchors.
However, if 9-inch holes were drilled, both layers of the No. 18 reinforcing bar would be cut. De' sign Change Authorization (DCA) No. 7401 was written for authorization to cut the uppermost No. 18 bar at
, only one rail, but it did not reference the authorization to cut the lowermost No. 18 bar. The DCA (No. 7041) also stated that the expan-sion bolts and baseplates could be moved in the east-west direction
- to avoid interference with the No. 11 reinforcement running in the north-south direction. The information described in DCA No. 7041 was substantiated by Gibbs & Hill calculations. The DCA approval was
-based on the understanding that only the uppermost No. 18 reinforce-ment would be cut. If the 10 holes were actually drilled 9 inches i deep, then the allegation that reinforcement was cut without proper j authorization may be valid.
The DCA indicated that the holes were drilled to accommodate 1/2-inch l Hilti bolts, which require a minimum embedment of 5-1/2 inches (as l noted in Fig. 39, Sh. 5 of 5, attached to DCA-7041). Since there was I no need to drill the holes deeper than 5-1/2 inches, the alleger may not be correct in stating that the holes were drilled 9 inches deep. Although the allegations discussed above, with the exception of AC-15 which requires further action, cannot be substantiated, the fact that such allegations were made indictted that there was no effective quality assurance program to oversee the issuance and use of diamond core drill bits. The TRT interviewed the individual concerned about the loan of rebar drills without proper documentation and the unauthorized cutting of rebar at. nonspecific locations to inform him of the TRT's finding. This individual did not agree with certain TRT findings. In K-89
e
- a. The allegations were not specific as to who made unauthorized cuts of 4
rebar or where the cuts took place.
- b. The number of unauthorized rebar cuts alleged, if true, would have an inconsequential effect on the safety of the structures.
However, the results of these evaluations will be further assessed as a part of the programmatic review concerning procedures addressed under QA/QC Category 6, "QC Inspection." Therefore, the final acceptability of these evaluations will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review wi11 be reported in a supplement to this SSER. Allegation AC-15 will remain open until the information requested of Texas Utilities Electric Company (TUEC) in " Actions Required" is provided! The TRT attempted to contact the' individuals who made allegations AC-18 and AC-15 to inform them of the TRT's findings. The individual who made allegation AC-18 will be informed of the TRT's findings by letter. One of the two individuals involved in allegation AC-15 cannot be located; the TRT is still attempting to contact the other. The TRT also contacted the individual who made allegations AC-13, AC-14, and AC-40 to discuss th.e TRT's findings pertaining to the concerns he raised in the first closure interview. An interview was arranged;.however, later the alleger indicated he did not want to meet with the TRT. A letter will be sent to him informing him of the TRT's findings.
- 6. Actions Required: TUEC shall provide:
- 1. Information to demonstrate that only the No. 18 reinforcing steel in l the first layer was cut, or
- 2. Design calculations to demonstrate that structural integrity is maintained if the No. 18 reinforcing steel on both the first and third layers was cut.
l l l K-91 l
- 1. Allegation Category: Civil and Structural 16, Excavation and Backfill
- 2. AllegationNumber$AQ-64
- 3. Characterization: It is alleged that overexcavation and improper fill under the Unit 1 Containment Building could invalidate the expected seis-mic response of the foundation due to the change in properties resulting from the removal of in-situ material.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT)
, did not initially attempt to contact the alleger because the allegation was sufficiently clear to allow the TRT to proceed with its investigation. During an investigation conducted in 1984, the NRC Office of Investigation (OI) interviewed the alleger (84-006, 3/7/84, A-7) and reference was made to overexcavation and improper repairs in the foundation rock for the Unit 1 Containment Building. The alleger stated that the excavation was erroneously made 6 to 8 feet too deep ard that upon realization of the error, the repair technique was simply to throw the loose rock back into the excavation and fill it in with concrete. The TRT reviewed NRC inspection reports, the FSAR, and the Atomic Safety and Licensing Board (ASLB) hearing transcript, where this concern was the subject of contention No. 7 and was admitted into the hearing on June 16, 1980. By order of March 5, 1982, the ASLB granted summary disposition of con-tention No. 7, based on the finding that no genuine issue as to any material fact was shown by any of the filings. The TRT also reviewed the affidavits and statements filed by TUEC and by the NRC in support of the motion for summary disposition. These documents adequately describe rock overbreak, accompanying fissures, and subsequent repairs. Affected areas were backfilled with concrete having a minimum compressive strength of 2,500 pounds per square inch at 28 days, or were grouted to maintain continuity of the competent rock in which fissures were identified. The TRT reviewed the procedures utilized to replace fractured rock with dental i concrete and to grout surrounding fissures and the accompanying compres-sive test results. The TRT found that FSAR figures 2.5.4-33a through 2.5.4-35 are maps of the excavation showing the location of fractures and the extent of dental concrete backfill. These figures showed that the area of overexcavation represented a small portion of the entire excavated area. FSAR figure 2.5.4-37, sheets 1 through 21, showed photographs of the excavated walls. The TRT interviewed the NRC inspector who was present during the excavation process and verified the conditions presented in the i FSAR. I The TRT independently evaluated the potential impact on the seismic re-sponse of the Unit 1 containment foundation due to the replacement of a limited amount of original rock with dental concrete from the standpoint of possible changes in foundation stiffness. Because of the facts that (a) the dental concrete's behavior, stiffness, and structural strength were essentially identical to those of the natural rock replaced at the
, K-93
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4
- 1. Allegation Category: Civil and Structural 17, Concrete Sampling
- 2. Allegation Number: AQC-45
- 3. Characterization: It is alleged that personnel produced incorrect readings on concrete batch plant scales by leaning on the wires connecting the weighing hoppers to the scales.
- 4. Assessment of Safety Significance: The NRC Tectnical Review Team (TRT) did not initially attempt to contact the alleget because the allegation was sufficiently clear to allow the TRT- to proceed with its investigation.
The allegation refers to "some type of sampling machine that tells whether there are good samples or bad samples in the concrete." The information is attributed to a friend of the alleger's who was an equipment operator at the concrete batch plant. NRC Inspection Reports 50-445/83-01 and 50/446/83-03 clarified the allegation, indicating that the equipment oper-ator is reported to have " charged that some personnel biased the operation of the concrete batch plant scales by leaning on the wires connecting the i scales to the sensors." , During the major construction phase of the project, the concrete plant
> consisted of two identical batching systems, one feeding a mixer drum and one batching directly into ready-mix trucks. Although the system with the self-contained mixer was removed prior to the TRT review, the other system was still in use at the time of the TRT review. The batching system in-cluded three mechanical lever dial scales, each controlled by a wire con-nected to the weighing hopper with which it is associated. Each wire entered the scale room over a pulley at the top of the room and ran down-ward vertically about 6 feet to the loading lever of the scale. The-scale dials faced the control room, from which they were visible through a large window. An electronic sensor connected to the scale provided a digital readout of each scale reading in the control room. It was possible to decrease the scale reading by entering the scale room and horizontally deflecting the vertical wires.
The TRT interviewed the concrete plant maintenance man who was present during most of the construction and who, at the time of the review, was also serving part time as operator because of the small amount of concrete production. He stated that the scale room was enclosed, well-illuminated, and provided with a large window so that all parts of the enclosure were visible from the control room, making it easier to prevent surreptitious tampering with the scale mechanisms. Thus, it would be obvious to everyone in the control room and to many outside the control room if anyone opened the scale room door, entered, and deflected the wires. The maintenance man was not aware that such an incident had ever occurred. I Since it was not possible to rule out such tampering completely, the TRT investigated the' potential consequences of such tampering. During a con-crete placement, a member of the TRT entered the scale room when all hoppers were loaded and deflected each scale wire as far as could be conveniently done by one person. The scale readings were affected as follows: K-95
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- 1. Allegation Category: Miscellaneous 1, Nuclear Fuel
- 2. Allegation Number: AM-2
- 3. Characterization: It is alleged that nuclear fuel wcs received prior to issuance of a special nuclear material (SNM) license.
- 4. Assessment of Safety Significance: The NRC Region IV (RIV) staff received this allegation by telephone in January 1983, from a member of the public.
The caller had friends working at Comanche Peak Steam Electric Station (CPSES) who claimed that fuel was recejved onsite before the NRC issued an SNM license. Despite requests from RIV for either more specific informa-tion about the allegation or the identity of those making the allegation, the NRC staff received no substantive information because the caller stated that the allegers feared they would lose their jobs. The NRC Technical Review Team (TRT) reviewed the SNM license for Texas Utilities Electric Company (TVEC), issued February 14, 1983, the " dummy fuel assembly" receipt package, and the first-fuel receipt package. During this review, the TRT found that TUEC received a " dummy fuel assembly" onsite on December 15, 1982, at 9:30 a.m., as noted on Form RF0-201-1, " Fuel Receiving Record - Shipment Report." Form RFO-201-1 also shows that the first-fuel shipment was received onsite on May 4, 1983, at
~
1:45 a.m. The TRT interviewed both the Radiation Protection Engineer (RPET and the Radiation Protection Supervisor (RPS) who stated that receipt of the
" dummy fuel assembly" was used as a training exercise to prepare for incoming shipments of fuel expected to arrive at CPSES. The RPS stated that when CPSES received the " dummy fuel assembly," fuel-receipt procedures were followed as if it were a "real case" shipment. Thus, it is conceiv-able that the alleger mistakenly assumed that the " dummy fuel assembly,"
which was received onsite prior to issuance of the SNM license, was actu-ally nuclear fuel.
- 5. Conclusion and Staff Positions: Based on a review of documents and forms, on interviews with the RPE and RPS, and on a review of an NRC Region IV
) interoffice memorandum assessing TUEC's program for receiving fuel, the TRT concludes that TUEC has an adequate program to receive fuel and did not actually receive fuel onsite prior to issuance of an SNM license on February 14, 1983. Accordingly, this allegation has neither safety l significance nor generic implications. The TRT was unable to learn the identity of the alleger during the inspec-tion; therefore, no followup interview with this alleger was possible. I
- 6. Actions Required: None.
l l l o K-97 l t
i
- 1. Allegation Category: Miscellaneous 2, Reactor Pressure Vessel
- 2. Allegation Numbers: AM-3 and AM-23b.
- 3. Characterization: Two allegations concerning the reactor pressure vessel
; (RPV) are characterized as follows:
- a. It is alleged that during hot functional testing (HFT) expansion caused the reactor pressure vessel reflective insulation (RPVRI) to come in contact with the concret,e biological shield wall (AM-3).
- b. It is also alls.ged that the Unit 1 RPV is located 3/16 inch to the west of the north-south centerline through the containment building (AM-23b).
t
- 4. Assessment of Safety Sionificance: The NRC Technical Review Team (TRT) reviewed the Final Safety Analysis Report (FSAR) for verification of RPV reflective insulation quality class; background material in the NRC Inspection Report 50-445/83-34; 50-446/83-18 and the affidavit of Doyle Hunnicutt clarifying this report; Brown & Root (B&R) construction opera-tion traveler 35-1195; and Westinghouse (W) Field Change Notices (FCNs)
- TBXM-10609, 10611, and 10612. The TRT also reviewed photographs of con-struction debris obstructions between the RPVRI and the biological shield and discussed all of these documents with the Texas Utilities Electric Company (TUEC) Mechanical Engineering Supervisor, the W project manager, and the TUEC heating, ventilation, and air conditioning system (HVAC) startup group leader. The TRT also reviewed documentation for hot func-
! tional test (HFT) No. ICP-PT-5502 for any information concerning the
- allegations, but found no specific notations about them.
The Comanche Peak Steam Electric Station (CPSES) FSAR, Volume XIV, Appendix 17-A, classifies the RPVRI as nonsafety-related. The support . ring (a support channel) for the RPVRI is part of the RPVRI; therefore, it l, too is considered to be nonsafety related. The only function of this sup-port channel is to support the refective insulation. It is not needed for [ the safe shutdown of the plant but simply insulates the vessel. In this ! case it assumed importance only because of its effect on adjacent safety- ! related structures or components. A Westinghouse FCN (TBXM-10609) dated 'p September 27, 1983, documented an unacceptable condition which was identi-fied during hot functional testing. Actual air temperatures during HFT were 288*F near the reactor vessel flange versus 150*F maximum in the cool-ing annulus between the RPVRI and the shield wall. Similarly unacceptable i temperatures were noted (150*F actual vs 135*F allowable) in the ex-core detector wel.ls. TUEC letter TX5054 reported extreme temperatures up to 314*F. Further examination by W personnel revealed that cooling air flow
- was restricted by the RPVRI support channel because of construction debris
- between the RPVRI support channel and the steel-lined concrett biological shield wall and because of restriction by the support channel. The debris i was removed by a remote technique and the gap was fiberoptically inspected.
Instead of the nominal design gap (cold) distance of 7/8 inch between the ! support channel and the shield wall, this inspection identified cold air gaps as follows: a 7/8-inch air gap extending one quarter around (90*) l t K-99
critical location factor, and alignment with the center of the Containment Building is of secondary importance. The Unit 1 RPV is alleged to be misaligned by 3/16-inch; however, opera-tion No. 7 of COT No. 35-1195 shows a setting tolerance of i 1/4 inch. The alleged 3/16-inch misalignment is within this tolerance. Attachment 35-1195-MCP-1 of COT No. 35-1195, " Alignment Location Record - Alignment Final Set," shows the maximum RPV deviation from the north-south and east-west centerlines of the Unit 1 Containment Building to be 0.003 inches which is within the specified tolerance. The TRT reviewed installation records and determined that the critical relationship of the NSSS compo-nents to each other, as well as to the Containment Building centerlines (N-S, E-W), was accurately maintained during installation of the reactor pressure vessel.
- 5. Conclusion and Staff Positions: Based on review of documentation and discussions, the TRT concludes that the RPVRI did make contact with con-struction debris, but did not contact the steel-lined concrete biological shield wall as specifically alleged. During fiberoptic inspection, TUEC personnel observed no visible damage to the reflective insulation, and all corrective modifications were accomplished and accepted in accordance with procedure MP 2.7.1-TBX-3 and FCNs TBXM-10609, 10611 and 10612.
The allegation, as specifically stated, cannot be substantiated, although it does have some merit because an unsatisfactory condition did exist in that the reflective insulation made contact with debris. However, this allegation has both safety significance and generic implications because of peripheral issues; i.e., failure to assure that proper design changes were communicated between organizations, failure to determine and report the un-derlying cause of a significant deficiency, and failure to ensure a proper gap between the support channel and shield wall when the vessel was set. The TRT also concludes that the RPV is set within the design location tolerance. Therefore, this allegation is not substantiated and has neither safety significance nor generic implications. The TRT is unable to interview the alleger to provide its findings and conclusions because the identity of the alleger is unlinown. The TRT could not identify the alleger responsible for allegation AM-3 because an anonymous alleger called the Dallas Times-Herald. The identity of the alleger of AM-23 is unknown because the person who received the allegation did not record the alleger's name and no longer remembers it.
- 6. Actions Required: TUEC shall:
- 1. Review their procedures for approval of design changes to nonnuclear
' safety-related equipment, such as the RPVRI, and make revisions as necessary to easure tt.at such design changes do not adversely affect safety-related systems. K-101 _._TZ_~_
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T
4
- 1. Allegation Cateaory: Miscellaneous 3, PSAR Errors j 2. Allegation Number: AM-4
.. 3. Characterization: It is alleged that Sections 10.2-11 and 10.2-12 of Volume VII of the Comanche Peak Preliminary Safety Analysis Report (PSAR) contain errors.
I
- t. 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) could not contact the alleger because the alleger's telephone number and -
i - address were unknown and relatives would not provide the TRT with this
- information. .
The TRT determined that this allegation refers specifically to PSAR Sec-tion 10.2-11 and 10.2-12; however, in the alleger's statement to the NRC Region IV Office of Investigation, the terminology "FSAR" (Final Safety Analysis Report) was frequently used when the term "PSAR" apparently was intended. The alleged errors refer to turbine missile energy calculations
; in PSAR Volume VII, Sections 10.2-11 and 10.2-12. Sections 10.2-11 and
[ 10.2-12 of the FSAR do not pertain to turbine missile energy. Th'e referenced PSAR sections contain a bounding design analysis calcu-l lationusedtodeterminethemaximumenergyava{1ableifagu'rbine
- failure occurs. Using the6equation, E = = w/g v the calculated
- j. value should be 18.27 x 10 ft-lbs;hoNver%mv, the PSAR erfoneo,usly stated
! the calculated value as 18.27 x 10 ft'lbs. It appears that the exponent
- was dropped because of a typographical error.
i The PSAR reflects the preliminary plant design, but it is the FSAR that i reflects the final plant design and safety analysis. Sections 10.2, 3.5.1.3, and 1.3 (Table 1.3-2, page 27) of the FSAR contain the final design and analysis for the turbine generator. These sections do not ! contain the alleged erroneous calculation. The FSAR refers to "Allis-4 Chalmers P.S. Inc., ER-503, ' Turbine Missile Analysis for 1800 t/ min NSTG with 44-inch Last Stage Blades,' July 1985." The analysis in this docu-ment is specifically applicable to the turbine generator which was actually purchased. The TRT reviewed the Allis-Chalmers analysis in the FSAR and determined that it corrected the turbine missile energy calculations in
, the PSAR. l The TRT found no errors in the Allis-Chalmers analysis of turbine. missile probability,~and concurred with the Texas Utilities Electric Company (TUEC) conclusion that even if missiles were generated, they would be contained by the low pressure turbine casing.
Although the alleged error appeared to be a typographical error, the TRT - randomly selected calculations from other FSAR sections and paragraphs (3.5A-1, 3.5-16A, and 10.3-5) for review and found no additional errors.
- 5. Conclusion and Staff Positions: The TRT reviewed the PSAR and determined that there was an error in the PSAR as alleged, but that it appeared to be' I
a typographical error. However, a review of the FSAR showed that the alleged error in the preliminary design calculation of the PSAR had been corrected in the FSAR. The NRC staff also randomly reviewed other calcu-lations in the FSAR and found that they were free of error. Additionally, K-103 l ..+, .- - e - ,.m-. .,-.,_.--.,..._w. -
- 1. Allegation Category: Miscellaneous 4, Radioactive Material Release
- 2. Allegation Number: AM-5
- 3. Characterization: It is alleged that someone " threw something radioactive in the lake," that is, in the Comanche Peak Reservoir, sometime between September and November 1978, and that this material may have been tritium.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) reviewed the background material contained in NRC Region IV Inspection Report No. 50-445/79-22. 50-446/79-21 which documents an inspection con-ducted September 7 and 12 and October 11, 16, 18 and 20, 1979, which followed up on allegations that appeared in the University of Texas at Arlington (UTA) newspaper, the " Shorthorn," on Wednesday, July 18, 1979.
This inspection concluded that "this allegation appears to' have no merit." The TRT also reviewed an NRC Office of Investigation transcript of inter-
- view No. 84-006, dated March 7,1984. From this document, the TRT learned that the alleger stated that he was looking through his turbine book and he saw that they used tritium gas to inspect for leaks. The alleger con-nected this with a story that another worker had told him, i.e., that one night the other worker was working down there when they threw something radioactive in the lake. The alleger in turn contacted the NRC to provide this information.
The TRT interviewed both the TUEC Radiation Protection Engineer (RPE) and the Radiation Protection Supervisor (RPS) about the allegation and about receipt of radioactive material at the Comanche Peak Steam Electric Station (CPSES), and also reviewed their radioactive source log. The RPS stated, and the radioactive source log documented, that first receipt of radioac-tive material at CPSES was on January 10, 1980, and that the quantity of strontium 90 and tritium received was exempt from licensing requirements. These small quantities were for laboratory use. The RPS also stated that although a leak-test procedure utilizing tritium could be employed for the turbine generato'r unit, the turbine generator at CPSES was hydrotested during hot functional testing, in lieu of the tritium leak-test procedure outlined in the manufacturer's instruction manual.
, The RPS stated that the first controlled shipment of tritium at CPSES was received on January 31, 1983. This shipment was authorized under a state
, of Texas radioactive material license (No. 5-2892) issued October 9, 1980, and the CPSES radioactive source log documented this receipt. The RPS also stated that an aliquot of this tritium standard was used to prepare a standard to calibrate the tritium monitors located on the turbine generator unit. The aliquot of tritium used to prepare the calibration standard was recovered, and CPSES verified its original radioactivity. The RPS stated that when the plant returns to hot furctional testing after fuel load, the primary coolant of the turbine generator will be " spiked" with this tritium. During operation of the turbine generator, a small amount of hydrogen gas will be extracted and measured by a tritium monitor. Thus, if there was a leak within the turbine generator cooling system, tritium would be detected in the hydrogen gas. l K-105 l l - .
- 1. Allegation Category: Miscellaneous 5, High Pressure Turbine
- 2. Allegation Number: AM-6
- 3. Characterization: It is alleged that cracks were observed in the lower casing of the high pressure (HP) turbine. The alleger did not specify the reactor unit where the cracks were alleged to exist.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) could nct contact the alleger because the alleger's telephone number and address were unknown and relatives would not provide the TRT with this information.
The TRT reviewed the applicable design, procurement, and vendor inspection information described in the Final Safety Analysis Report (FSAR), Sec-tion 10.2, " Turbine Generator"; Allis-Chalmers, Turbine Description, No. 1-1-0200-7163; TUGC0 Purchase. Order, No. CP-0003; and, Vendor Surveil-lance Report No. 294. The HP turbine was supplied by Allis-Chalmers Power Systems, Inc. , and is classified as nonsafety related. The HP casings were fabricated in Japan by a casting process; however, they were subsequently shipped to the Mulheim plant in Germany where manufacturing was completed. The vendor surveillance referenced above was performed in Germany by a Gibbs & Hill (G&H) QC engineer on October 3-6, 1977. Records indicate that the engineer visually examined the Unit 1 HP turbine, witnessed hydrostatic testing, and reviewed the iupporting documentation, which included nondestructive examination (NDE) records. The vendor surveillance report on the Unit 1 HP turbine concluded that inspection and testing showed the casing to be satisfactory. On August 4, 1984, the TRT interviewed both the Texas Utilities Electric Company (TUEC) supervisor of turbine construction and the TUEC lead startup engineer. They stated that they observed no cracks in the casing. In addition, the lead startup engineer was present when the Unit 1 turbine was rolled to synchronous speed during testing, and he indicated that no casing problems (including casing leaks) were observed and that only an
; insignificant diaphram leak was detected during testing. The TRT also i reviewed test documentation which showed acceptable results.
The TRT inspected the outside of the upper and lower casing of the HP tur-bine for Unit 2 and found no cracks. They did not inspect the casing of i the Unit 1 HP turbine because it is nonsafety related and is wrapped with ! insulation. Since independent vendor inspection and test records, and TUEC observation and testing, revealed no unacceptable conditions, the TRT did not request removal of the insulation or did not perform further casing inspections of the Unit 1 turbine. I l 5. Conclusion and Staff Positions: The TRT determined that TUEC and vendor ! personnel performed visual inspections, witnessed hydrostatic and startup testing on the HP turbine for Unit 1, and found no unacceptable conditions. The TRT also inspected the outer casing of the Unit 2 turbine and found no cracks. Moreover, this equipment is classified as nonsafety related and is not needed for the safe shutdown of the reactor. e K-107 i
\
1' ,
$ 1. Allegation Category: Miscellaneous No. 6, Pressurizer Area Piping
- 2. Allegation Number: AM-7
- 3. Characterization: It is alleged that an 18-inch section was cut from a
} prefabricated pipe "in the pressurizer area."
- 4. Assessment of Safety Significance: The NRC Technicc1 Review Team (TRT) t attempted to contact the alleger to determine the exact location of the cut piping because it was not specifically identified in the allegation.
- However, the alleger's telephone number and address were unknown, and rela-tives of the alleger would not provide the NRC staff with this information.
Due to the size of the piping (18 inches), and because the alleger stated that the " cut pipe was close to the pressurizer," the staff assumed that l' the cut section of piping was located in the pressurizer system. Based on a review of a Final Safety Analysis Report (FSAR) flow diagram (" Reactor i Coolant System," Section 5.1) and a walkdown inspection of the pressurizer j area, the NRC staff determined that there is no 18-inch line in the pres- ! surizer system and immediate area and that the 12-inch piping that runs ! from near the top of the pressurizer (at approximately the 907-foot eleva-
- tion) to the pressurizer relief tank (at the 834-foot elevation) most I closely fit the description given in the allegation.
o l The TRT also reviewed the isometric drawings for Unit 1 and Unit 2. These , show the as-built configuration of the four 6-inch lines which run from
- the pressurizer to a common 12-inch line which, in turn, runs to the pres-j surizer relief tank. In this run of piping are both smaller piping and a l 3-inch line which runs from the 12-inch line to the safety relief valves,
, and which ties into piping in the residual heat removal system (RHR) suc-
! tion (Trains A and B) and for chemical and volume control system (CVCS) l seal return and letdown. All modifications to the pipe lines identified
- in the isemetric drawings, including the trimming of pipe for fitup, were i recorded on component modification cards (CMCs). Trimming modifications j of 7/8-inch and 1-1/2-inch made on 12-inch piping were recorded on CMCs j
61551 and 47943R1, The TRT inspected and counted the number of welds in { all of the above piping runs, then compared that number with both the num-ber shown on the as-built isometric drawings and the number on the CMCs. All three numbers were the same, and the NRC staff found no evidence of unauthorized work in the piping system. Because the 12-inch piping runs are nonsafety related and not essential for ! safe shutdown of the plant, the quality assurance requirements of 10 CFR Part i 50 Appendix B are not applicable; however, Texas Utilities Electric Company I (TUEC) technicians monitored the piping installation in accordance with CP-CPM-6.9, " Welding and Related Processes." The TRT selected and reviewed two welding records and two test reports which confirmed that appropriate i procedures were used to control welding. Again, the TRT found no evidence of unauthorized work, i ) . ! I l K-109 l* - i emg%ip ww*qv---m-ye.-w-- -w ---,g. 9,yswas------,.w7+---7-ym- +vwe---g-y ww-r -sp.,
F
- 1. Allegation Category: Miscellaneous 7, Condenser i
; 2. Allegation Number: AM-8,~AM-9 and AM-10 4
3 3. Characterization: It is alleged that: (a) the Unit 1 main condenser tubes ! were beaten with air and sledge hammers, were split during belling and j flaring, and were improperly rolled; (b) the wrong type condenser for the i steam generator was used; (c) the tube support sheets had holes that were j misaligned by 3/8 inch; and, (d) the turbine-to-condenser tubing was mis-aligned and jacked into alignment, causing stress.
- 4. Assassment of Safety Stanificance: The NRC Technical Review Team (TRT)
- . interviewed the alleger on August 24, 1984, to gather additional infor-nation regarding his concerns.
j A Region IV letter to TUEC identified the potential safety concerns asso-i ciated with this allegation and asked Texas Utilities Electric Company (TUEC) to respond in writing. In a letter dated July 9, 1984, TUEC out-L~ lined their monitoring procedures for condensers during and after con-struction activities, as well as.the various tests that they had conducted. This letter stated that testing demonstrated that no leakage of lake water , into the steam generator occurred because of differential pressure that is maintained during operation. It also states that the main condenser [ and circulating water systems are not safety related and therefore not sub-j ject to the quality assurance requirement of 10 CFR Part 50, Appendix B. i Regardless, construction and fabricati'on were done in accordance with sound l engineering and construction practices.
! In assessing this allegation, the TRT reviewed NRC Office of Investigations
- interview 84-006; the RIV letter and TUEC response referenced above; TUEC l office memoranda 50-81051, 5U-81081, and SU-81134; and Condenser Retest i No. 1 CP-AT-27-01. These documents show that TUEC was well aware of and j was controlling and correcting problems associated with the condenser
- fabrication and operation.
l The TRT attempted to visually inspect the condenser internals, but this l was not possible because of the congestion caused by the tube bundles, the internal piping, and the bracing. Therefore, the inspection was limited to using a hand-held light and observing the bundles from the manway j opening. Observation was difficult because of the distance between the j manway and the tube bundles and because of surface rust; however, the
- l. limited inspection detected no irregularities.
The TRT interviewed three employees who had direct knowledge of the work , j on the main condenser tubes. With respect to part (a) of this allegation, l 1.e., use of hammers and splitting tubes, a Brown & Root (B&R) Mi11 wright ! Superintendent stated that hammers were used on the condenser tubes, but only after the ,sta11ation of a special tool to protect the tube ends. He also stated that this is a standard practice in condenser assembly. Concerning part (c) of this allegation, the M111 wright Superintendent
- disclaimed any knowledge of misaligned tube sheet holes.
The TRT also interviewed a TUEC Operations Results Engineering Supervisor i concerning the availability of documents such as inspection reports, l 1 K-111
f I 2 I the TUEC Operations Maintenance Foreman showed that rewelding began on i November 13, 1978, and was completed on November 17, 1978, and that move-ment was controlled to a maximum of 0.023 inches (0.584 mm). All welding i was done to B&R weld procedure (No. 10046) and to Westinghouse directions for skip (intermittent) welding. Considering these welding controls, I there was no evidence that jacking occurred or that any undue or undesir-4 able stress was introduced into the welded joints. Following welding, the i only NDEs performed (other than visual inspections) were hydrostatic and i vacuum tests which were successfully completed. 4 The TRT also reviewed seven condenser hydrostatic test packages for the
- results of tests conducted in accordance with the G&H test procedure ,
i (Specification 2323-M-23). These tests spanned the period from December 1980 to April 1984. The TRT found that retests of the shell were made i following design changes. The last of these retests, No. ICO-0200E, was i performed in September 1983, and involved sodium tracer injection and
- sampling at points through the condenser wall. In all cases the tests were successfully completed. In May 1984, TUEC performed a vacuum and j water box priming retest (Procedure ICP-AT-27-01-RT-1) to again verify j that the main condenser could be evacuated by the vacuum pumps and hold
- $ vacuum for 1 hour. The system test engineer witnessed and accepted this sucessfully completed test. The lead startup engineer, the manager of plant operations, the TUSI nuclear engineering manager, and the manager of j nuclear operations also reviewed and accepted the test results. The TRT's review of these documents indicated that there was no evidence of an over-stress problem with either the expansion joint or the piping' connection.
i 5. Conclusion and Staff Positions: The TRT determined that this unit is classified as nonsafety related and is not essential for the safe shutdown of the plant, and confirmed that when TUEC began the tube-rolling procedure, I they experienced some fabrication problems; however, these problems appear to have been solved. The TRT also confirmed that TUEC plans to retube the condenser with titanium tubes to improve its design and operation. l The TRT found no evidence that holes in the tube support sheets were mis-aligned by 3/8-inch, nor did they find evidence that the turbine to con- - denser was misaligned to the point that excessive stress was introduced. Th'e conclusive evidence is that the condenser was constructed following i approved construction and testing procedures and, as such, will perform < > its design function. ! The TRT concludes that this allegation was not substantiated. However,
- it was true that fabrication problems occurred and that condenser redesign '
i (tube material changes) and misalignment occurred, but not as alleged. Accordingly, this allegation has neither safety significance nor generic ! implications. l The TRT attempted to present its finding and conclusions to the alleger ' in a follow-up interview, but the alleger could not be located. The TRT was unable to find either a telephone number of an address for the alleger. 2 A letter will be sent to the intervenor of record, outlining the resolu-tion of the alleger's concerns. A
- 6. Actions Required: None.
K-113 m . . _ . . ..
s
- 1. Allegation Category: Miscellaneous 8, Damaged Component Cooling Water Tank Supports
- 2. Allegation Number: AM-12
- 3. Characterization: It is alleged that during the installation of Unit 1 component cooling water (CCW) surge tank, the anchor bolts were damaged.
- 4, Assessment of Safety Significance: The NRC Technical Review Team inter-viewed the alleger on August 24, 1984, to gather additional information - regarding his concerns. The TRT reviewed the background material which alleged that the anchor bolts were beaten sideways with a hammer to make them line up with holes in the plate and were overtorqued to the point that they stretched. The CCW surge tank, part of the component cooling water system, is listed in Final Safety Analysis Report (FSAR) Table 17A-1 as Safety Class 3, ASME III Code Class 3, and Seismic Category I. The centerline of the CCW surge tank for Unit 1 is at an elevation of 889 feet, 6 inches in the Auxiliary Build-ing; the baseplate attaches to bolts embedded in concrete which interface at an elevation of 895 feet, 6 inches. In assessing this allegation, the TRT reviewed Texas Utilities Services Inc. (TUSI) Drawing N-2640-359. The TRT also visually inspected the installed CCW surge tank and saw no stripped threads or bent or cracked bolts. Two nuts and one washer were present on each bolt. The CCW surge tank has 10 bolts on each end which support the tank. A review of the installation documentation revealed that 5 out of the 20 concrete anchor bolts were misaligned. A TUSI letter (CPP-00825), dated March 2, 1979, documents the need for modifying the baseplate holes. This letter, which describes the misalignment, indicates that the bolts were not installed as required by the specification and drawing. As a result, i component modification card (CMC) No. 4263 was approved June 8, 1979. The TRT found no nonconformance reports (NCRs) in the quality assurance (QA) records vault. The misalignment may have caused installation problems; however, if any damage occurred at that time it was not documented. Traveler No. ME78-108-1101 and a traveler revision record sheet document the installation of the CCW surge tank. This traveler was initiated on October 10, 1978, and the tank was placed on June 13, 1979, when the Millwright Supervisor signed the traveler. This traveler failed to give instructions for tightening nuts on anchor bolts as required by Procedure 35-1195-MCP-1, Revision 2, paragraphs 4.1.10 and 4.1.11. Both the quality control (QC) inspector and the Millwright Supervisor signed the traveler on June 18, 1979, indicating that the tank was level and located as recorded on the ai-built drawing. The TRT interviewed the Millwright Supervisor and the QA/QC and engineer-ing personnel listed on the traveler and learned that although they had
' direct knowledge of work activities on a day-to-day basis, they had no knowledge of bent or cracked bolts or of damaged threads. The engineer K-115
l l
- 1. Allegation Category: Miscellaneous 9, Hayward Tyler Pump Deficiencies
- 2. Allegation Number: AM-13
- 3. Characterization: It is alleged that Comanche Peak Steam Electric Station (CPSES) has pumps in safety systems manufactured by Hayward Tyler Pump Co.
(HTPC) that may have unidentified deficiencies because of a poor quality assurance (QA) program by HTPC.
- 4. Assessment of Safety significance: The NRC Technical Review Team (TRT) interviewed the alleger on August 24, 1984, to gather additional infor- ,
mation regarding his concerns. . The NRC Technical Review Team (TRT) learned that in late 1981, the NRC received allegations that upper management of Hayward Tyler Pump Company of Burlington, Vermont failed to support the quality assurance program. In 1982, the NRC Region IV staff inspected the HTPC QA program after receiving these allegations. The investigation established that signifi-cant deficiencies existed in the implementation of HTPC's QA program from 1977 through 1981. As a result, the NRC issued a report and a Notice of Nonconformance on December 22, 1982c As a result of the above findings, the NRC also issued IE Bulletin No. 83-05 to licensees and applicants on May 13, 1983. This bulletin addressed HTPC's failure to effectively implement its QA program, and required that specific actions be taken by the holders of operating licenses and construction permits who were using or planning to use HTPC pumps in safety systems. IE Bulletin 83-05 requirements for successful pump operability applied to both the original pump assembly and to reassembled pumps which use spare parts. The bulletin required the following actions from NRC applicants and licensees: to provide NRC with the number of HTPC pumps and their service appli-cation, l to provide NRC with a summary of the in-service test requirements for the affected pumps and spare parts, to conduct a pump performance test by running the pump continuously for a minimum of 48 hours without maintenance or repair, with the test incorporating specific criteria provided by HTPC, l to provide NRC with the results of the required ASME Code hydrostatic pressure test, to implement HTPC recommendations with respect to fitup and dimensional considerations during installation of spare parts, to conduct a pump performance test when spare parts are installed, l unless it could be demonstrated that the spare part in question j would not affect any parameters that are measured, and 4 K-117
=
- 1. Allegation Category: Miscellaneous 10, Damaged Diesel Generators
- 2. Allegation Number: AM-14
- 3. Characterization: It is alleged that the Unit 1, Train A, diesel generator was damaged in May 1982, because of improper handling practices.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) tried to contact the alleger during its inspection to learn more dete.ils about the allegation. The TRT reviewed all nonconformance reports (NCRs) issued in May 1982, that pertained to the er.iergency diesel generators '
(EDGs). Of the 11 NCRs reviewed, 4 (E-82-005335, E-82-00560, E-32-006065, and E-82-004795) documented equipment or instrument damage; however, the necessary corrective action was taken and the NCRs were closed appropriately. The TRT determined that it was unlikely that any damage that occurred in May 1982, could now affect the operability of the EDGs in light of the extensive EDG inspection and testing that took place in 1984. This inspec-tion and testing was the result of generic EDG problems which Transamerica Delaval, Inc. (TDI) and owners of TDI's EDGs identified to the NRC. On August 12, 1983, the main crankshaft on one of the three EDGs at Shore-I ham Nuclear Power Station broke into two pieces during a load test. TDI issued several 10 CFR Part 21 reports that reflected a variety of major and minor defects. These defects included cracks in piston skirts, push rod cracks, governor drive coupling failures, potential failures in fuel lines, and dimensional problems with component fasteners and dowel pins. Although there are some design differences between the EDGs at Comanche Peak Steam Electric Station (CPSES) and those at other plants, the identified defects were generic in nature. During the evaluation of the failure and repairs of the Shoreham EDGs, information related to the operating history of TDI engines and the QA program of the manufacturer was identified which called into question the reliability of all TDI diesels. As a result of the evaluation and its generic implications, representatives from affected nuclear power plants [ formed an " Owners' Group" to investigate all aspects of the quality and reliability of TDI-supplied EDGs. The Owners' Group developed a generic inspection program. This program ( addressed the specific concerns brought about by defects reported to the NRC by owners of TDI EDGs, 10 CFR Part 21 reports from TDI, and other areas of concern in order to develop adequate confidence in these EDGs. Texas Utilities Electric Company (TUEC) also expanded some inspections of the TDI EDGs by increasing sample sizes, inspecting other areas on their own initiative, and inspecting both of the Unit 1 EDGs between February and June 1984. The inspections included disassembly aad nondestructive examinations of parts using methods such as radiography, liquid penetrant testing, magnetic particle testing, visual inspections and measurements, eddy current testing, ultrasonic testing, and metal comparator testing. , TUEC then transmitted the results of their inspections to the Owners' i Group for evaluation and incorporation into the recertification process. l l l K-119
4
, l l
- 1. Allegation Category: Miscellaneous 11, Polar Crane Shimming
; 2. Allegation Number: AM-15, AM-16
- 3. Characterization: It is alleged that the shimming of the Unit 1 polar crane rail system supports was improper and that the polar crane system is
- improperly installed.
- 4. Assessment of Safety Sionificance: The NRC Technical Review Team (TRT) tried to contact the alleger during its inspection to learn more details about the allegation. On August 14, 1984, two TRT members visually
- inspected the shims from the polar crane. During the first 180* rotation 1 of the crane, the TRT members stood on the platform above the operator's booth to view the radial restraint brackets and the seismic restraint
) brackets. Several brackets at different locations appeared to have gaps in excess of 1/16 inch. However, this only confirmed what had been
; previously observed by an NRC Region IV Resident Inspector and was docu- '
mented in NRC Inspection Report (IR) No. 50-445/84-08, which required .
! corrective action which was not yet completed. - ! The TRT then moved below the operator's booth to view the polar crane rail
- system from another vantage point. The TRT observed the shims used for
; shimming 28 crane girder to girder-support brackets. During this 180*
rotation, the TRT observed large gaps, particularly on the inside edge l (looking from the inside of the Containment Building to the outside). ll l The TRT met with the Texas Utilities Electric Company (TUEC) project civil 1 engineer, the Brown & Root (B&R) project control manager, the B&R subcon-tracts supervisor, and a representative from Chicago Bridge and Iron I
! (CB&I) to determine the gap-tolerance specification between bearing plate
! "A" (Dwg. 2323-S1-0515, Revision 4) and the girder to girder-support
- bracket. Neither Gibbs & Hill (G&H) specification SS-14 nor the Crane Manufacturers Association of America Manual (CMAA-70) addressed this issue.
The meeting failed to produce a specific answer; however, copies of two ' i letters related to the issues were provided. The first, a B&R letter (No. BRF-7404), dated November 8, 1977, contained the as-built measure- ; ments of gaps at all shim locations and a request for G&H to evaluate this information and provide direction. At 28 locations, the as-built drawings showed gaps that ranged up to 0.581 inch. In the second letter, G&H
- (GHF-2207, dated November 28, 1977) responded as follows
l Girder Seat Connections These seated connections will not require shimming since the area in i bearing is at least the width of the bottom flange of the crane girder. < The gap dimensions indicated in the Brown & Root survey exist only at the extreme edges of plate A, Section 3-3, Dwg. 2323-S1-0515, Revi-
- sion 4.
! The TRT noted that the bottom flange of the girder referenced in the G&H letter (the bearing surface) is 20 inches wide. i
~
On August 30, 1984, an NRC inspector, accompanied by a TUEC quality con- { trol (QC) inspector, inspected the 28 crane girder to girder-support
. K-121 1
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ngny
i i e 4 Note: The gaps in the seismic restraints were the subject of NRC Inspec-tion Reports 50-445/82-11, 50-446/82-10, and 50-445/84-08; violations were issued in each report. Although these matters may have been evaluated and a response made to the referenced violations, TUEC shall consider this matter as a part of the inspection of the polar crane system.
> 0 0
0 0 K-123
4
- 1. Allegation Category Miscellaneous 12, Welding of Lifting Lugs onto Tornado Missile Barrier Doors
- 2. Allegation Number: AM-17
- 3. Characterization: It is alleged that deficient welds on a missile barrier door were accepted.
- 4. Assessment of Safety Sionificance: The NRC Technical Review Team (TRT) tried unsuccessfully to contact the alleger during its inspection to learn more details about the allegation. The TRT. identified the location of the doors from information provided by Texas Utilities Electric Company (TUEC) quality assurance personnel who were present at the time of the allegation, and from a review of the Atomic Safety and Licensing Board (ASLB) hearing record on intimidation and harassment. The door referred to in this alle-gation is the tornado missile barrier located at ground level on the west side of the Unit 1 diesel generator room. The alleged deficient welds are 18 double grooved welds that attach the lifting lugs to the three missile barriers. These welds were terminated by wrapping the weld around the end of the lug, a practice questioned by the alleger.
The TRT learned that the alleger mistakenly believed that neither the welds which were made nor the wraparounds terminating the welds were allowed by the weld symbol specified on the drawing. The alleger thought that a lifting lug which was welded to the flat side of the missle barrier steel door (using a double grove T-weld joint) should have been indicated with a weld symbol showing a double groove on each side of the lifting lug and with a weld symbol at the end of the lug showing a fillet weld where the runoff occurred. The Brown & Root (B&R) inspectors who.actually performed the inspections did not interpret it as the alleger did and accepted the 18 disputed welds. The TRT reviewed the B&R inspection reports, which indicated that welds were performed in accordance with Welding Procedure Specification (WPS) 10046, Rev. 9, and with American Welding Society (AWS) code requirements. The TRT also reviewed the construction traveler and the inspection reports for the shop fabrication and field-fitting of the missile barriers and found that, although rework occurred, no rework was done on the lifting lug welds. The TRT reviewed the inspection procedure, weld procedure, and inspection reports referred to in the traveler for applicability and compliance with AWS Code 01-1, Sections 4.6.1 and 4:6.2, to determine if the code allowed wrapping the weld around the end of the lug. These sections of the code state: 4.6.1 Groove welds shall be terminated at the ends of a joint in a manner that will ensure sound welds. Whenever possible, this shall be done by the use of extension bars or run-off plates. K-125 9 -
i t
- 1. Allecation Category: Miscellaneous 13, Welding of Pipe Supports in Safeguards Tunnel
. 2. Allegation Number: AH-18
- 3. Characterization: It is alleged that the tube steel used to fabricate supports by welding it to baseplates in the Unit 1 safeguards "796 yard tunnel" was cut at the wrong angle, resulting in too large a gap between the tube and baseplate.
- 4. Assessment of Safety Stanificance: The NRC Technical Review Team (TRT) visually observed the entire "796 yard tunnel" to determine how many sup-ports were used, ar.d found there were several hundred tube-steel-to-baseplate weldments/ installations. A member of the TRT also attempted to contact the alleger ta obtain additional information to determine the approximate location, size, and configuration of the subject tube-steel-to-baseplate weldments, because without this information indiscriminate destructive test-ing of the installed supports would be necessary in order to identify the location of the alleged gap. However, the alleger refused to communicate with the TRT.
The TRT visually observed the "796 yard tunnel" to identify any condition that would show improper installation or welding on the 796-foot, 6-inch elevation, and the three, short 800-foot-elevation tunnels. Subsequently, the TRT randomly selected typical pipe supports and five hanger inspection reports (DD-1-16-025-Y33R, 00-1-16-024-Y33R, SI-1-031-041-532K, SW-1-17-716-Y33K, and AF-1-002-033-Y33K) and reviewed them to determine if the documentation of the inspections required during installation was cor-rect. The inspection reports indicated that the supports were correctly installed in accordance with Brown & Root procedures CP-CPM-6.9E, " Pipe Fabrication and Installation," Revision 7; CP-CPM-6.9F, " Fabrication and Installation of Component Supports," Revision 0; and CP-CPM-7.1G, " Piping Supports," Revision 0. The TRT requested that a Brown & Root inspector take copies of the inspec-tion reports for the five pipe hangers to the field and repeat those steps in the inspection which could be repeated; however, no root gaps could be inspected because all welding had been completed. Inspection included dimensional checks, measuremen) of welds, checks for proper anchoring, and visual inspection of welds. The TRT found no major discrepancies between the inspection reports and the field conditions. The recheck of the five inspection packages also showed that the as-installed and as-inspected conditions agreed, which indicated that both workmanship and inspection on the five hangers were adequate. Most of the rework required by the non-conformance reports (NCRs) found in the inspection packages consisted of filling in undersized fillet welds, although some rework was initiated,by design change authorization (DCA). , The TRT learned that from July to September 1984, the NRC Region IV (R1V) inspectors reviewed a portion of the Unit 1 auxiliary feedwater system while performing inspection 50-445/84-26. The RIV inspectors paid specific attention to two water lines which were connected to the condensate water storage tank (CP-AFATCS-01) and were located in the safeguards tunnel. K-127 O
i t 1 i The TRT was unable to provide the above findings and conclusions to the alleger because the alleger could not be located. Several attempts were made by letter and telephone to locate this alleger; however these attempts were unsuccessful.
- 6. Actions Required: None.
S e O D 1 K-129 e
- 1. Allegation Category: Miscellaneous No. 14, Posting of NRC Form-3
- 2. Allegation Number: AM-19
- 3. Characterization: It is alleged that posting requirements for NRC Form-3 were not met during 1977-1982.
~
- 4. Assessment of Safety Sionificance: The NRC Technical Review Team (TRT) reviewed the allegation, which was made in NRC Office of Investigation (01) document A4-83-005 dated May 20, 1984, and found no need to contact the alleger to further clarify the allegation.
The TRT reviewed the deposition given by Robert R. Taylor dated July 17, 1984, in which this NRC inspector stated that Texas Utilities Electric Company (TVEC) had posted a memorandum or letter in 1978 (about 6 months before Taylor became the resident inspector at the plant). This letter invited any site employee to contact the NRC if they had concerns about the quality of the construction of the plant. The NRC Region IV telephone number was the point of contact. This letter was the forerunner of the NRC Form-3 which became a posting requirement in October 1982. Mr. Taylor could not recall if the Form-3 was posted in October 1982, but had pre-pared a report in January 1983, which documented the posting. The Atomic Safety and Licensing Board (ASLB) deposition of C. Tedder, H. Hollis, C. daker, and M. Hall, dated July 18, 1984, stated that NRC Form-3 had been posted from October 1982 until the present. The bulletin boards were periodically checked to assure proper posting. The TRT reviewed 10 CFR Part 50 dating back to 1976, and learned that the 10 CFR Part 50.7 requirement for posting the NRC Form-3 around sites under construction was not effective until October 12, 1982. The TRT observed the locations of the 12 Comanche Peak Steam Electric Station (CPSES) pro-ject bulletin boards currently in use, as well as 5 additional bulletin boards in various work spaces. The TRT determined that Form-3 was properly posted on all bulletin boards. The TRT interviewed the TUEC Radiation Protection Engineer (RPE), who currently is responsible for maintaining the CPSES Unit 1 bulletin boards, and the TUEC administrative and control supervisor, who is now responsible for maintaining the 12 CPSES project bulletin boards. The RPE stated that TUEC designates official bulletin boards in work areas and other assembly areas and reviews them periodically to ensure compli-ance with posting requirements. The TUEC administrative and control super-visor also stated that the CPSES project bulletin boards are reviewed perio-dically to ensure compliance with posting requirements and that locations for the bulletin boards may change as construction progresses. Management has not formally assigned responsibility in writing for establishing bulle-tin board locations or for maintenance and periodic review; the responsi-bility is informally assumed. The TRT also telephoned the Texas Utilities Service, Inc. (TUSI) personnel manager who was responsible for maintenance of the bulletin boards during the September 1982 to October 1983 period. The TUSI personnel manager stated that approximately five bulletin boards were in place and that an K-131
- 1. A11enation Category: Miscellaneous 15, Drug Abuse
- 2. Allegation Number: AM-21
- 3. Characterization: In a letter dated March 7, 1984, it is alleged that there was widespread drug use and abuse at the Comanche Peak Steam Electric Station (CPSES) and that management did not give proper attention to the alleged problem.
- 4. Assessment of Safety Sionificance: The NRC Technical Review Team (TRT) reviewed background material and an NRC Region IV report of inspection pursuant to temporary instruction (TI 2596/1) that documented discussions held with representatives of Texas Utilities Electric Company (TUEC) in early 1983. This report described TUEC's company policy on use or posses-sion of drugs and alcohol, employee assistance programs, background checks /
psychological tests, supervisory and employee awareness of drug / alcohol problems, and a drug / alcohol abuse detection program. During its review, the TRT learned that a drug abuse prevention program had been in effect at CPSES since 1974, when work on the project began, and that it included TUEC policy statements which emphasized that any employee possessing drugs or alcohol on company property was subject to immediate discharge. As a part of the review, the TRT interviewed managers, staff, and technicians affiliated with TUEC and Brown & Root (B&R). The staff also interviewed medical laboratory personnel and law enforcement officials in the CPSES area. The topics discussed in these interviews included the following:
- The methodology used in conducting drug investigations.
- The techniques used during pre-employment background investigations.
- The program for supervisors to ensure that they recognized employees with potential problems.
- The employee assistance program for permanent TUEC employees.
The TRT determined through interviews and a review of personnel depart-ment practices and records that B&R corporate policy required prospective employees to take a physical examination to satisfy insurance or workmen's compensation requirements; however, the urinalysis given in this physical examination did not include an analysis for drugs. , The TRT also learned that as a precondition of employment, both at TUEC and at Burns International (the CPSES physical security contractor), any employee requiring unescorted access to secured areas had to provide a urine specimen which was analyzed for a wide range of drugs. Prospective TUEC and security contractor employees were alao subje.cted to an in-depth background investigation. TUEC's physical security plan commits them to mandatory screening and investigation-of potential employees when the reactor becomes operational; however, TUEC elected to put these policies into effect prior to reactor operation. K-133
l might have been inadequate. This program involved determining the total number of inspections for each of the inspectors and selecting a statisti-cal sampling plan from MIL-STD-1050, " Sampling Procedures and Tables for Inspection by Attributes." The sampling plan taken from MIL-STD-105D provided for a General Inspection, Level-II, single, normal inspection with an acceptable quality level (AQL) of 4.0 considered to.be adequate. The TRT reviewed the results of TUEC's reinspection program and found that the reinspected items / components were randomly selected and that a valid sample was reinspected by B&R inspectors. On August 31, 1984, the TUEC QA staf.f engineer stated that no significant deficiencies had been identified during their reinspection effort; however, eight minor deficiencies were referred to engineering in TUEC IOM QA-0047, dated September 21, 1984. IOM QA-0047 also included a request for an evaluation of the safety impli-cations of these minor deficiencies had they gone undetected. (The final sign off of this NCR was not completed as of January 31, 1985, pending TUEC engineering and legal review.) Following TUEC's reinspection program, the TRT randomly selected five per-cent of the TUEC sample to verify the adequacy of the reinspection. (The B&R inspectors involved in drug-related activities were identified as A, B, C, D, E, F, and G.) The TRT also included the work of the two inspec-tors whose work had not been reinspected by TUEC/B&R personnel. One B&R inspector's work was not included in this sample because it pertained to coatings, an area which was extensively inspected and evaluated by the TRT Coatings Group. Based on this sample, the TRT determined that the inspections were adequate. In addition, the TRT found no items / components that were deficient. The TUEC QA manager stated that although some craft personnel had been involved in drug abuse, their work was not reinspected because they were not responsible for final acceptance of their own work, but relied on in-process inspections and a final acceptance inspection made by B&R inspec-tors who were not involved in the drug-related incident. In addition, an ANI inspected work done by craft personnel when ASME work was involved. The TRT evaluated this position and reasoned that the sample of all inspec-tor's work would also include a sample of craft personnel work, and if no significant deficiencies were found, their justification would be accepted.
- 5. Conclusion and Staff Position: Based on the above review, the TRT con-cludes that TUEC had performed an investigation and identified B&R per-sonnel implicated by their refusal to take polygraph tests and their subsequent termination of employment. Although this allegation had potential safety significance and generic implications, TUEC wrote a nonconformance report which identified all work performed by the impli-cated B&R inspectors and reinspected by different inspectors. The rein-spection identified only minor deficiencies that have been referred to engineering for final evaluation and correction. This allegation appears to have some substance.
With respect to management, the TRT concluded that TUEC and site contrac-
, tor management and supervision had implemented strong measures to prevent K-135
E i I
- 1. Allecation Category: Miscellaneous 16, Heating, Ventilating, and Air j Conditioning System <
- 2. Allecation Number: AM-22
! 3. Characterization: It is alleged that Texas Utilities Electric Company l (TUEC) has not analyzed the heating, ventilating, and air conditioning
- system (HVAC) supports for seismic loads; that all HVAC components and
. supports inside the Containment Building were not properly considered in-
- regard to their treatment as missiles; that the HVAC system is not properly supported; and, that HVAC failure during a postulated accident would allow the temperatures to rise to an unacceptable level inside the Containment l
Building. j 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) found no need to contact the alleger to further clarify the allegation. i The TRT reviewed the Comanche Peak Steam Electric Station (CPSES) Final Safety Analysis Report (FSAR) to identify the HVAC's design and quality { assurance requirements. FSAR Volume IV, Section 3.2, " Classification of l' Structures, Components and Systems," states that part of the containment ventilation system is seismic Category I; however, FSAR Volume XIV, Sec-tion 17.0, Appendix 17A, " List of Quality Assured Items," states that the - I containment ventilation system (which contains eight subsystems / components) is seismic Category II and nonsafety related with the exception of the containment purge exhaust ductwork, supports, debris screen, and isolation
! valves, which are seismic Category I. Only the isolation' valves, which
! are safety and code class 2, are safety related and seismic Category I. 4 l The TRT determined that the entire containment ventilation system is nonsafety related, except for the isolation valves referenced above. None i of these nonsafety-related systems is necessary for the safe shutdown of ' l the reactor or to prevent or mitigate the consequences of accidents or
- malfunctions in the reactor coolant pressure boundary. All components, *
. except the containment purge exhaust, which are inside the Containment Building are seismic Category II, and need not operate during a safe shut- .
down earthquake (SSE) but simply are components which are not aMowed t.o , i fall and damage an essential safety-related system. Therefore, the HVAC , I ductwork is not required to remove heat from containment. The system that ' removes heat from Containment Building is the containment spray system, r i which does not depend on HVAC duct work or HVAC supports. The allegation , that temperatures would rise to an unacceptable level because of an ; inoperative HVAC is incorrect. : The containment purge exhaust is classified as nonsafety related and i seismic Category 1, which means it is designed to continue operating during , l a SSE; however, this ductwork system is not essential for the safe shutdown , of the plant. The containment isolation valves close on a signal of high radiation to prevent a release to the environment, as specified by ; 10 CFR Part 20. This system is not used to remove heat from the Contain-ment Building either, as previously discussed. This is a containment spray system function so that the alleged high temperatures could not be caused by inoperative containment exhaust HVAC ductwork and supports. t
~
K-137
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5
- 1. Allegation Category: Miscellaneous 17, Damage to Upper Internals
- 2. Allegation Number: AM-24
- 3. Characterization: It is alleged that damage occurred to the 15-foot by 2 1/2-inch stainless steel bars (subsequently determined by the NRC staff to be thermocouple columns) located in the reactor vessel upper internal structures in the Unit 1 Reactor Building at Comanche Peak Steam Electric Station (CPSES). The alleger's concern is that approximately 1 foot from the top of the stainless steel bars, two of then were bent when they were struck by either a fork lift or a crane. The alleger contends that a rope pulled by a crane was then placed around the stainless steel bars and pulled in order to straighten them. It is further alleged thct no docu-mentation was ever completed to show that this damage occurred.
- 4. Assessment of Safety Significance: NRC Region IV (RIV) inspected this allegation and documented the results in Inspection Report 50-445/84-08, 50-446/84-04 (July 26, 1984). Prior to this inspection (March 1984), the RIV discussed this allegation with the alleger by telephone. There was also an NRC Office of Investigation (01) inquiry on this matter.
The NRC Technical Review Team (TRT) reviewed OI report QA-84-016 dated April 11, 1984, the notes from the telephone conversation with the alleger, and NRC followup on Inspection Report 40-445/84-08, 50-446/84-04. In addi-tion, the TRT visually inspected the upper internal structures of the reac-tor vessel and reviewed a computerized index of CPSES documentation, i.e. , nonconformance reports (NCRs) and/or procedures related to damage or prob-lems in the upper internals or reactor vessel head areas. Two documents, Westinghouse Field Deficiency Report (FDR) TBXM-10285 and Brown & Root (B&R) NCR M-11438, indicated that on October 14, 1983, the refueling crane (a bridge crane that straddles the refueling cavity) was moved without crane interlocks or a " flagman," a condition which resulted in a bent thermocouple column. Although interlocks are normally used in such a case, at the time the alleged damage occurred no specific procedure was in effect that required use of interlocks. The TRT learned that when the alleged damage occurred, the upper core assembly was mounted on extension legs and was stored in its designated location in the refueling cavity. The extension legs elevated the upper internals so that the thermocouple column was in the refueling crane's normal path. Each of the four thermocouple columns (tubes) is approxi-mately 17 feet long and provides support for the incore thermocouple tubing located between the upper core internals and the reactor vessel head; the bottom of each thermocouple column is attached to the upper core assembly. The thermocouples in these columns are chromel-alumel wires that are threaded into guide tubes which penetrate the reactor vessel head through seal assemblies and terminate at the top end of the fuel assemblies. Thermocouple readings are monitored by a computer, with backup readout pro-vided by a precision indicator from the incore instrumentation, even if the computer is not in service. These thermocouples are not required for safety. (See Final Safety Analysis Report [FSAR], Section 7.7 " Control Systems Not Required for Safety.") K-139 ,
On November 1, 1984, the TRT provided the above findings and conclusions to the alleger. T.he alleger stated that his questions or concerns were answered and he had no further concerns.
- 6. Act.fons Required: None.
=
9 4 + i e I ' K-141 y +*-*W- '
- 1. Allegation Category: Miscellaneous 18 Broken Internal Wires in Polar Crane Festooned Cable and Crane Movement Interference
- 2. Allegation Number: AM-25
- 3. Characterization: It is alleged that internal wires were broken in the polar crane festooned cables and that the polar crane hit unspecified hangers while operating.
- 4. Assessment of Safety Significance: The alleger was interviewed by the NRC Technical Review Team (TRT) on August 3, 1984, to obtain additional infor-mation regarding the allegaticn. On August 30, 1984, the TRT and a Texas Utilities Electric Company (TUEC) quality control (QC) inspector visually examined the festooned cables. There was no visible damage on any of the cables. In addition, a review of preoperational inspection megger test data sheets revealed that all tests were satisfactory.
The TRT visually inspected the polar crane during three rotations to deter-mine if there were any interferences between the crane and supports or other installed items, and noted no interferences. It is possible that the alleger was referring to the problem of the uplift lugs striking the crane girder stiffener plates, which is described in nonconformance report (NCR) M-81-00054; however, this problem was resolved in accordance with DCA 11311, Rev. 1. On September 12, 1984, the TRT, accomp'anied by a Brown & Root (B&R) QC electrical inspector and electrician, opened the two electrical junction boxes that feed the festoons on the polar crane walkway. The inspector visually inspected all of the wires in both boxes and found no broken or non-terminated wires. The TRT asked the operator of the crane about pro-blems with the crane, specifically asking if the limit switches cut out properly. The crane operator then demonstrated the operation of the bridge crane, running it until the limit switch cut out and a signal light indi-cated that it cut out. Again, he stated that there were no problems with the crane. L The TRT determined from a B&R QC inspection that no records or nonconform-ance reports existed which may have documented the alleged defective j festooned cables because they were classified as nonsafety related. The TRT l found only the records for megger testing which were previously discussed.
- 5. Conclusion and Staff Positions: The TRT found no damaged festooned cables.
However, the polar crane uplift lugs did strike the crane girder stiffener plates and this had potential safety significance and generic implications. The TRT did find an NCR documenting the damaged plates corrective action, l and the TRT verified that corrective acticn was taken, i.e., the crane now operates without such interference. Based on a review of applicable.docu-mentation, examination of the polar crane cables and wiring on the polar t crane walkway, and interviews with the crane operator, the TRT concludes ! that this allegation was substantiated; however, appropriate corrective action was taken. Accordingly, this allegation has neither safety signifi-cance nor generic implications. i i ' K-143
- 1. Allegation Category: Miscellaneous 19, Chloride Contamination of Radwaste System Piping
- 2. Allegation Number: AM-30
- 3. Characterizstion: It is alleged that workers habitually urinated on stainless steel pipe located in the radwaste system.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) found no need to contact the alleger to further clarify the allegation.
The TRT reviewed background information pertinent to this allegation and searched the Texas Utilities Electric Company's (TUEC) quality assurance records relating to piping cleanliness. Norconformance report (NCR) M-82-00305 described an instance where piping in the radwaste area (above the waste monitor tanks in Room 2) was " contaminated by unknown liquid substances of unspecified chemical composition." The NCR stated that hold tags were applied and lines 2 WP-X-218-151-R5 and 2 WP-X-208-151-R5, located in the Unit 1 Auxiliary Building (elevation 790 feet of Gibbs & Hill drawing 2323-Al-0507 Revision 9), were subsequently cleaned and swipe-tested according to Procedure QI-QP-11.1-65 to assure that the sur-faces were free of chlorides and fluorides. TUEC verified corrective action and closed this NCR on May 27, 1982. The TRT discussed this NCR with TUEC quality ana engineering personnel, who stated that the NCR quoted a report made by a QC inspector who wit-nessed a worker urinating on the piping. TUEC personnel further stated that they knew of no other similar instances; however, all safety-related stainless steel piping surfaces (outside) are routinely cleaned prior to final turnover. The safety significance related to chlorides, a chemical present in human urine, on stainless steel surfaces depends on the service conditions and residual stresses that may be present. If excessive stress (near the yield strength) and chloride contamination are present, the alloy may fail because of stress corrosion cracking. Since the alloy is expected to operate with a design load applied, it is necessary to ensure that chlorides are not present. In this case, the piping was cleaned to remove any chlorides that could have been deposited by urine. Thermal insulation is applied after cleaning and this protects safety-related piping (necessary for safe shut-down) from further contamination. The radwaste piping above the radwaste i tanks is nonsafety related and is not needed for the safe shutdown of the j plant. On August 29, 1984, the TRT inspected the radwaste areas (Rooms 179, 184, and 185) and found them locked and access to them controlled. Housekeeping
~
appeared to be excellent, and the' TRT detected no odors which might indicate that the area was further contaminated. The number of craft personnel who work in the Unit 1 buildings has been limited for several months, as com-i pared to earlier periods, cecause this unit is virtually completed. ! Because limited work is in progress and personnel access controls are in ! place, it appears unlikely that other similar incidents occurred after cleaning. . l l . i I K-145 i -- - - - . . . -. .
- 1. Allegation Category: Miscellaneous 20, No Procedures or Guidance Provided for Rigging and Handling Large Components / Equipment
- 2. Allegation Number: AM-23(a)
- 3. Characterization: An NRC Region IV Resident Inspector identified a vio-lation as a result of a discussion with a craft person who stated that he had not received instructions about how to rig and handle a large motor-operated valve.
- 4. Assessment of Safety Significance.- The NRC Technical Review Team (TRT)
Tound no need to contact the alleger to further clarify the alleg,ation. The TRT reviewed NRC Inspection Report 50-445/79-27, 50-446/79-26 and its corresponding Notice of Violation (NOV). The TRT also reviewed the Texas Utilities Electric Company (TUEC) response to these documents (TXX-3080, dated December 18,1979), which stated that the subject valve was not mis-handled, nor was it damaged. The engineering organization had not, how-ever, reviewed specific vendor rigging or handling recommendations or noted the procedures for loads exceeding 2000 pounds. An NRC followup inspec-tion verified that Brown & Root (B&R) Procedures CP-CPM-6.3, 35-1195-CCP-24, 35-1195-ACP-3, and QI-QAP-13.1-1 were reviewed by TUEC and revised appro-priately. NRC Inspection Report 50-445/80-18, 50-446/80-18 (dated September 19,1980) documented corrective action during the followup inspection. The TRT interviewed TUEC's Rigging Craft Superintendent, Assistant Mechan-ical Superintendent, and Senior Staff Engineer. They stated that the revised procedures (specifically, CCP-2A, Revision 4, " Rigging"; CP-CPM-6.3, Revision 10, " Preparation, Approval, and Control of Operation Travelers"; and, CP-CPM-6.9, Revision 2, " General Piping Procedure") adequately con-trolled heavy lifts of equipment and components. Nonconformance report (NCR) M-2128 documented the problem which was identified as a viola-tion, and the appropriate site personnel reviewed the NRC inspection report and concurred with the corrective action. In addition, the TRT independ-ently reviewed the revised procedures for the control of heavy lifts of equipment and found the control of rigging and handling to be acceptable for loads less than or exceeding 2000 pounds. l
- 5. Conclusion and Staff Positions: The TRT determined that Region IV (RIV) f l confirmed that the craftperson's stated need for better instructions was correct and confirmed follcwup inspection by the RIV inspector to verify that corrective action was accomplished in accordance with TUEC letter TXX-3080 (December 18,1979). The TRT concludes that the failure to pro-vide proper instructions for rigging and handling heavy loads is safety significant and has generic implications; however, corrective action was taken. No evidence of further inadequacies in this area was found; con-sequently the allegation requires no further action.
l l The TRT tried to provide the above findings and conclusions to the i alleger; however, the allsger's identification is unknown.
- 6. Actions Required: None.
i K-147
Attachment 3 y
/ '%,
4- ' UNITED STATES NUCLEAR HEGULATORY COMMISSION j .f.[.,i WASHINGTON. D. C. 20555
*A';jl
/
t$E? 1.8 004 Dockets: 50-445 50-446 Texas Utilities Electric Company Attn: M. D. Spence, President, TUGC0
, Skyway Tower ,
400 North Olive Street
- Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
SUBJECT:
COMANCHE PEAK REVIEW On July 9, 1984, the staff began an intensive onsite effort designed to complete a portion of the reviews necessary for the staff to reach its i decision regarding the licensing of Comanche Peak Unit 1. The onsite effort covered a number of areas, including allegations of improper construction practices at the facility. The NRC assembled a Technical Review Team (TRT) responsible for evaluating most of the technical issues at Comanche Peak, including allegations. The TRT has recently identified a number of items that have potential safety implications for which we require additional information. These items are listed in the enclosure to this letter. Further background information regarding these' issues will be published in a Supplement to a Safety Evaluation Report (SSER), which will document the overall TRT's assessment i of the significance of the issues examined. The items in the enclosure to this letter, which are in the general areas of electrical / instrumentation, civil / structural and test programs, cover only a portion of the TRT's effort. The TRT evaluation of items in the areas of mechanical, QA/QC, and coatings, and its consideration of the programmatic implications of these findings, are still is progress. A summary of these issues will be provided to you at a later date. 7 You are requested to submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified. This program plan and its implemen-tation will be evaluated by the staff before NRC considers the issuance of , an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic implic-ations on safety-related systems, programs, or areas. The collective significance of these deficiencies should also be addressed. Your program plan should also include the proposed TUGC0 action to assure that such problems will be precluded from occurring in the future. 1 K-149 -
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q -- -
COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels & Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas' 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallas, Texas 75201 B. R. Clements Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B.'81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq. P. O. Box 355 1200 New Hampshire Avenue, N. W. Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol .Staticn Government Accountabi.11ty Project Austin, Texas 78711 1901 Que Street, N. W. Washington, D. C. 20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq. Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Teras 75224 San Francisco, Californit 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq. Trial Lawyers for Public Justice CYGNA 101 California Street 2000 P. Street, N. W. San Francisco, California 94111 Suite 611 Washington, D. C. 20036 K-151
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3
' Panel CP1-ECPRCB-14, Cable E0139880 Panel CP1-ECPRTC-16 Cable E0110040 Panel CP1-ECPRTC-16, Cable E0118262 Panel CP1-ECPRTC-27, Cable EG104796 Panel CPX-ECPRCV-01, Cable EG021856 Panel CP1-ECPRCB-02, Cable NK139853 (nonsafety)
Accordingly, TUEC shall reinspect all safety-related and associated terminations in the control room panels and in the tennination cabinets in the cable spreading room to verify that their locations are accurately depicted on drawings. Should the results of this reinspection reveal an unacceptable level of nonconformance to drawings, the scope of this reinspection effort shall be expanded to include all safety-related and associated terminations at CPSES. .- 5. The TRT found cases where nonconformance reports (NCRs) concerning vendor-installed terminal lugs in GE motor control centers had been improperly closed. Exa~mples are NCR Nos. E-84-01066 through NCR E284-01076, inclusive. Accordingly, TUEC shall reevaluate and redisposition all NCRs related to vendor-installed terminal lugs in GE motor control centers. ! b. Electrical Equipment Separation The TRT reviewed the separation criteria between separate cables, trays and conduits in the main control room and cable spreading room in. Unit 1, and the compatibility of the electrical erection specifications with regulatory requirements. The TRT reviewed documentation and inspected random samples of separation between safety-related cables, trays and conduits and between them and nonsa.fety-related cables, trays and conduits. I
+In numerous cases, safety-related cables within flexible 1.
conduits inside main control room panels dic not meet minimum seperation requirements. Examples are as follows: Panel CP1-EC DRCB-02 Panel CP1-EC-PRCB-07 , Panel CP1-EC-PRCP-06 ( Panel CP1-EC-PRCB-08 Panel CP1-EC-PRCB-09 Accordingly TUEC shall reinspect all panels at CPSES, in ' addition to those in the main control room for Unit 1, that l contain redundant safety-related cables within conduits, or I safety and non-safety related cables within conduits, and either correct each violation of the separation criteria, or I . K-153
Table 1 Examples of Cases of Safety or Nonsafety-Related Cables In Contact With Other Safety-Related Cables Within Conduits in Control Room Panels
- 1. Control Panel CPI-EC-PRCB Containment Spray System .
Cable No. Train Related Instrument EG139373 een IJndetemined E0139010 W(orange)) A Undetemined .
- 2. Control Panel CP1-EC-PRCB Reactor Control System Cable No. Train Related Instrument EG139383 NT9reen) Reactor manual trip switch E0139311 A(orange) Undetennined
' 3. Control Panel CP1-EC-PRCP Chemical 3 Volume Control System Cable No. Train Related Instrument EG139335 Wgreen) LCV-112C E0139301 A(orange) Undetermined
- 4. Control Panel CP1-EC-PRCB Auxiliary Feedwater Control System Cable No. Train Related Instrument E0139753 A (orange) TK-2453A E0139754 A orange) FK-2453B E0139756 B green) FK-2454A EG139288 8 green) FK-24548
'l K-155
,-. -m- w n
i One case of no documentation to waive the remaining 2 months of the required 1 year experience. One case where a QC technician had not passed the required color vision examination administered by a
- professional eye specialist. A makeup test using colored pencils was administered by a QC~~ supervisor, was passed, and then a waiver was given.
Two cases where the experience requirements to become a Level 1 technician were only marginally met.
. One case of no documentation in the training and certification files substantiating that the person met the experience requirements.
Accordingly, TUEC shall review all the electrical QC inspector training, qualification, certification and recertification files against the project requirements and provide the infomation in such a fom that each requirtment is clearly shown to have been met by each inspector. If an inspector is found to not meet the training, qualification, certification, or recertification requirements, TUEC shall then review the records to detemine the adequacy of inspections made by the unqualified individuals and provide a statement on the impact of the deficiencies noted on the safety of the project.
- 2. The TRT found a lack of guidelines and procedural requirements for the testing and certifying of electrical QC inspectors. Specifically, it was found that:
No time limit or additional training requirements existed between a failed test and retest. I No controls existed to assure that the same test would not be given if an individual previously failed that test. No consistency existed in test scoring. No guidelines or procedures were available to control the disqualification of questions from the test. No program was available for establishing new tests (except when procedures changed). The same tests had been utilized for the last 2 years. Accordingly, TUEC shall develop a testing program for electrical j QC inspectors which provides adequate administrative guidelines, l procedural requirements and test flexibility to assure that suitable proficiency is achieved and maintained. l l K-157 i
v i
?
x E g interviews with fifteen individuals. Of these, only the alleger and one other individual stated they thought that - falsification-occurred, but they did not know when or by whom. The ' TRT also reviewed slump and air entrainment test results of concrete , placed Februaryduring)the 1977 and did period theany not find alleger was variation apparent employed in the (January 1976 to - uniformity of the parameters for concrete placed during this period. Although the uniformity of the concrete placed appears to minimize the likelihood that low concrete strengths were obtained, other allegations were raised concerning the falsification of records associated with slump and air content tests. The Region IV staff addressed these allegations by assuming that concrete strength test results were adequate. Furthermore, a number of other allegations dealing with concrete placement problems (such as deficient aggregate grading and concrete in the mixer too long) were also resolved by assuming that concrete strength test results were adequate. The TRT agrees with Region IV that, while the preponderance of evidence suggests that falsification of results did not take place, the matter cannot be resolved completely on the basis of concrete strength test results, especially if there is ar.y doubt about whether they may have been falsified. Due to the importance of the concrete strength test results, the TRT believes that additional action by TUEC is necessary to provide confirmatory evidence that the reported concrete strength test results are indeed representative of the strength of the concrete installed in the Category I concrete structures. Accordingly TUEC shall determine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a program to assure acceptable concrete strength. The program shall , include tests such as the use of random Schmidt hanner tests on the concrete in areas where safety is critical. The program shall include a comparison of the results with the results of tests per-formed on concrete of the same design strength in areas where the strength of the concrete is not questioned, to determine if any significant variance in strength occurs. TUEC shall submit the program for performing these tests to the NPC for raview and approval prior to performing the tests.
- c. Maintenance of Af r Gap Between Concrete Structures The TRT investigated the requirements to maintain an air gap between concrete structures. Based on the review of available inspection reports and related documents, on field observations, and on discussions with TUEL engineers, the TRT cannot cetermine whether an adequate air gap has been provided between concrete structures. Field investigations by B&R QC inspectors indicated unsatisfactory conditions due to the presence of debri,s in the air 9
K-159 ,
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- ,-n- --- -- - - . _ _ . _ _ - , _ _ . _ _ _ _ __ ____ _ _ _ _ _ _ __ _
- 1. Heating, Ventilating and Air Conditioning - Seismic Category I
- 2. Safety-Related Conduits - Seismic Category I
- 3. Nonsafety-Related Conduits - Seismic Category II
- 4. Lighting Fixtures - Seismic Category II
- 5. Sloping Suspended Drywall Ceiling - Non-Seismic
( 6. Acoustical Suspended Ceiling - Non-Seismic 7; Lowered Susper.ded ceiling - Non-Seismic According to Regulatory Guide 1.29 and FSAR Section 3.7B.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failurs would not adversely affect the functions of safety-related components or cause injury to operators. For the nonseismic items (other than the sloping suspended drywall ceiling), and for nonsafety-related conduits whose diameter is 2 inches or less, the TRT could find no evidence that the possible effects of a failure of these items had been considered. In addition, the TRT determined that calculations for seismic Category II components (e.g., lighting fixtures) and the calculations for the sloping suspended drywall ceiling did not adequately reflect the rotational interaction with the nonseismic items, nor were the fundamental frequencies of the supported masses detennined to assess the influence of the seismic response spectrum at the control room ceiling elevation would have on the seismic response of the ceiling elements. Accordingly, TUEC shall provide:
- 1. The results of seismic analysis which demonstrate that the i
' nonseismic items in the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. An evaluation of seismic design adequacy of support systems for the Ifghting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems.
- 3. Verification that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
- 4. The results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
K-161
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l (TDRs) that were generated as a result of test deficiencies a found prior to and during HFT.
- 1. Chapter 14 of the FSAR and Regulatory Guide 1.68 provide requirements for the conduct of preoperational testing.
In reviewing test data packages, the TRT found that certain test objectives were not met. It appears that the Joint Test Group approved incomplete data packages for at least three preoperational hot functinal tests. These were: Test Procedure Defigiency . ICP-PT-02-12 " Bus ~ Because acceptable voltages Voltage and Load Survey" could not be achieved with the
- specified transformer taps, they were changed. A subsequent engineering evaluation required returning to the original taps, but no retest was performed.
1CP-PT-34-05, " Steam Level detectors 1-LT-517, 518 Generator Narrow Range and 529 were rep, laced with Level Verification" temporary equipment of 3 design that was different from that which was to be eventually installed ICP-PT-55-05 Level detector 1-LT-461 appeared
" Pressurizer Level to be out of calibration during the Control" test and was replaced after the test.
The retest approved by the JTG was a cold calibration rather than a test consistent with the original test objective, which was to obtain satisfactory data under hot conditions. Accerdingly, TUEC shall review all complete precperational test data packages to ensure there are no other instances where test objectives were not met, or prerequisite conditions were not satisfied. The three items identified by tie TRT shall be included, along with appropriate justification, in the test deferral packages presented to the NRC. L l . 4 K-163
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l Apparently after repairing leaks found during the first two attempts..the third attempt at a CILRT was successful. It was successfully completed after three electrical penetrations were isolated because the leakage through them could not be stopped. Though the leaks were subsequently repaired and individually tested with satisfactory results, NRC approval was not obtained to perfom the CILRT with these penetrations isolated. In addition, leak rate calculations were performed using ANSI /ANS 56.8, which is neither endorsed by the NRC nor in accordance with FSAR commitments. , Accordingly TUEC shall identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rather than ANSI N45.4-1972. Additionally, TUEC shall identify to NRC all other deviations from FSAR comitments.
- c. Prerequisite Testing The TRT reviewed FSAR comitments, startup administrative procedures, prerequisite test records, craft personnel qualification records, and discussed them with startup and craft management personnel. The TRT also observed test support craft personnel at work and interviewed soite of them to gain familiarity with their attitudes and capabilities.
The review of test records revealed that craft personnel were signing to verify initial conditions for tests in violation of startup Administrative Procedure-21, entitled: " Conduct of Testing" (CP-SAP-21). This procedure requires this function to be performed i , by System Test Engineers (STE). Startup management had issued a memorandum improperly authorizing craft personnel to perfom these i verifications on selected tests. Accordingly TUEC shall rescind the startap menoran fum (STM-81084), l which was issued in conflict with CP-SAP-21, and ensure that no other memoranda were issued which are in conflict with approved procedures. i d. Preoperational Testing The TRT assessed the preoperational test program by reviewing , administrative procedures, interviewing startup personnel, and examir.fng test records, schedules, system assignments, subsystem definition packages, and the master data base. l Problems found with test data are addressed in section III.a of this l enclosure. The TRT also found that STEs were not being prcvided with I current design information on a routine, controlled basis, ano had to update thcir own material when they considered it appropriate. l Accordingly TUEC shall establish measures to provide greater .
- assurance that STEs and other re'sponsible personnel are provided with l
current controlled design documents and change hotices. K-165 l
Attachment 4 1954 Occket Nos.: 50-445 0QT5
. and 50-446 Texas Utilities Electric Company Attn: M. D. Spence, President, TUGC0 Skyway Tower
, 400 North Olive Street Lock Box 81 - Dallas, Texas 75201
Dear Mr. Spence:
Subject:
September 18', 1984 Letter, D. G. Eisenhut to M. D. Spence, Re: Comanche Peak Review During our meeting on September 18, 1984 at Bethesda, Maryland, we discussed the technical issues regarding Comanche Peak which the NRC Technical Review Team identified as having potential safety implications and thus requiring additional .information. The subject letter listing these items and the information that we requested were provided to you during that meeting. We have since discovered some typographical errors in the Enclosure to the September 18, 1984 letter and provided Mr. John Merritt of your staff with a marked-up copy of that letter on September 21, 1984. Enclosed for your information is an errata to the letter. Sincerely, Original signed by Darre11 G. Eisenhut Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation '
Enclosure:
l As stated cc w/ enclosure: See next page K-167
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CMANCHE PEAK Mr. M. D. Spence Pres'ident Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dsiles, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector /Ccmanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory Washington, D. C. 20036 , Comission
. P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge s. Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000
~
Mr. Mcmer C. Schmidt Arlington, Texas 76011
. Manager - Nuclear Services
. Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallas, Texas 75201 B. R. Clements Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. 0. Box 355 1200 New Hampshire Avenue, N. W. Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. . Billie Pirner Garde Environmental Protection Division Citizens Clinic Director l P. O. Box 12548, Capitol Station Government Accountability Project ! Austin, Texas 78711 1901 Que Street, N. W. Washington, D. C. 20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq. Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams ' Anthony Z. Rcisman, Esq. CYGNA Trial Lawyers for Public Justice 101 Califcenia Street 2000 P. Street, N. W. San Francisco, California 94111 . Suite 611 Washington, D. C. 20026 K-169
# "49 Attachment 5
# \ UNITED STATES
! NUCLEAR REGULATORY COMMISSION .
h WASHINGTON, D. C. 20555
%*****) NOV 2 9 E84 Docket Nos.: 50-445 and 50-446 Mr. M. D. Spence President Texas Utilities Generating Company 400 North Olive Street Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
Subject:
Comanche Peak Review On July 9, 1984, the staff began an intensive onsite effort to complete a por-tion of the reviews necessary for the staff to reach its decision regarding the licensing of Comanche Peak, Unit 1. The onsite effort covered a number of areas, including allegations of improper construction practices at the facility. On September 18, 1984, the NRC met with you and other Texas Utilities Electric Company representatives to provide you with a number of technical issues in the electrical / instrumentation, civil / structural, and test program areas having potential safety implications. The issues discussed constitute a portion of the technical issues and allegations being evaluated by the Technical Review Team (TRT). The activities of the TRT have progressed to the point where it is appropriate to provide yoJ with a status of additional items under review and to request additional information. These items, in the coatings, mechanical, and miscel-laneous areas, are listed in the enclosure to this letter. Further background information regarding these issues will be published in a Supplement to a Safety
- Evaluation Report (SSER), which will document the TRT's overall assessment of l the significance of the issues examined.
l ! The items in the enclosure to this letter cover only a pertion of the TRT's effort. The TRT's ongoing evaluation, QA/QC review and conversatiens with allegers may reveal additional items in the coatings, mechanical, and mis-cellaneous areas for which additional requests for information ray be appro-priate. Also,.the TRT evaluation of QA/QC issues, and its consideration of the.programmitic impi.ications of these findings, are still in progress. A summary of these issues will be provided to you at a later date. l You are requested to .eubmit additional information to the NRC, in writing, in-cluding a program and schedule for completing a detailed and thorough assess-ment of the issues identified in the enclosure to this letter. This program plan and its implementation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic K-171
COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Canpany 400 N. Olive St., L.B. 81
- 2. Dallas, Texas 75201
. cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop. Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. cI/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. Robert D. Martin 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive i Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services
! Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallas, Texas 75201 B. R. Clements Vice President Nuclear j Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. O. Box 355 1200 New Hampshire Avenue, N. W. Pittsburgh, Pennsylvania 15230 Suite 420 . Washington, D. C! 20036 Renea Hicks, Esq. J l Assistant Attorney General Ms. Billie Pirne~r Garde i Environmental Protection Division Citizens Clinic Director l P. O. Box 12548, Capitol Station Government Accountability Project . Austin, Texas 78711 1901 Que Street, N. W.
' Washington, D. C. 20009 Mrs. Juanita Ellis, President
! Citizens Association for Sound David R. Pigott, Esq. ! Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street
, Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq.
i CYGNA Trial Lawyers for Public Justice < 101' California Street 2000 P. Street, N. W. San Francisco, California- 94111 Suite 611
- Washington, D. C. 20036-K-173
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i REQUEST FOR ADDITIONAL INFORMATION IV. Protective Coatings Area
- a. Surveillance and Test Program for Coatings
- The. protective coatings Technical Review Team (TRT) reviewed the backfit program, design basis accident qualifications, traceability, application ~
and repair procedures, training, coating exempt log and dispositioning of non-conformance reports. Concurrently, the staff is evaluating the effects on containment emergency sump performance of paint and insulation debris. The results of the two concurrent reviews will be combined in one supple-mental safety evaluation which is scheduled to be issued by January 1985. Actions required for resolution of protective coatings issues will be delineated in the supplement. V. Mechanical Area
- a. Inspection for Certain Types of Skewed Welds in NF Supports The TRT investigated inspection procedures of Brown & Root (B&R) for welds in pipe supports designed to ASME III Code, Subsection NF. The TRT found that no fil.let weld inspection criteria existed for certain types of skewed welds. By definition, skewed welds are those welds joining (1) two non-perpendicular or non-colinear structural members, or (2) two members with curved surfaces or curved cross sections, such as a pipe stanchion (a sec-tion of pipe used as a structural member) welded to another pipe stanchion or to a curved pipe pad. Notice that for type (2), the effect of curva-ture at the weld connection induces skewed considerations, even though the two joining members are physically perpendicular. The B&R weld inspection
- procedures CP-QAP-12.1 and QI-QAP-11.1-28 for NF supports have addressed type (1) skewed welds; however, the TRT found that QI-QAP-11.1-28 did not include weld inspection criteria for type (2) skewed welds. Although the TRT was told by B&R personnel that procedure QI-QAP-11.1-26 for piping
! weld inspection.was used, since such weld connections were similar in con-figuration to a pressure boundary stanchion attachment weld, no evidence documenting the use of this inspection procedure was provided to the TRT. According to records reviewed by the TRT, these welds were actually cate-gorized as "all other welds" rather than " skewed welds" on the required QC checklist. Instead of using fillet weld gauges for measuring the size of nonskewed welds, welders were supposed to use a straight edge and a steel-scale for measurement of a type (2) skewed weld, as described in QI-QAP-11.1-28. In addition, due to the variable profile along its curved ' weld connection, the weld size saould have been measured at several dif-ferent locations. The lack of inspection criteria and lack of verification of proper inspection procedu,res being conducted for type (2) skewed welds are a violation of ASME Code for NF supports committed to by TUEC in FSAR Section 5.2.1 and a violation of Criterion XVII in Appendix B of 10 CFR 50. K-175
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(1) replace shortened bolts with bolts of proper length, or provide analysis to justify the adequacy of shortened bolts as installed; and, (2) provide justification or propose measures to ensure that no similar concern" exists for bolting.
- c. Design Consideration for Piaing Systems Between Seismic Category I and Non-Seismic Category I luildings In April 1984 the Comanche Peak Special Review Team (SRT), formed and coor-dinated between NRR, IE and Region II and IV, performed . limited review of Comanche Peak. The TRT, in reviewing the SRT findings in' the area of piping design considerations, has discovered that piping systems, such as Main Steam, Auxiliary Steam and Feedwater, are routed from the Electrical 4 Control Building (seismic category I) to the Turbine Building (non-seisaic category I) without any isolation. To be acceptable, each seismic cate-gory I piping system should be isolated from any non-seismic category I
; piping system by separation, barrier or constraint.
If isolation is not feasible, then the effect on the seismic category I piping of the failure in the non-seismic category I piping must be considered (CPSES FSAR 3.78.3-13.1). For CPSES, FSAR section 3.78.2.8 establishes that the Turbine Building is a non-seismic category I structure and failure is postulated during the seismic (SSE) event. The effect of Turbine Building failure on any non-isolated piping routed through the Turbine Building from any seismic i category I building must be considered. In addition, for non-seismic category I piping connected to Seismic Category I piping, the dynamic effects of the non-seismic category I piping must be considered in the seismic design of the seismic category I piping and supports, unless TUEC can show that the dynamic effects of the non-seismic category I piping are isolated by anchors or restraints. The anchors or restraints used for isolation purposes must be designed to ' withstand the combined loading imposed by both the seismic category I and l non-seismic category ! piping. . Accordin51 y, TUEC sball provide analysis and documentation that the piping systems routed ftom seismic category I co non-seismic category I buildings ' l meet the stated FSAR criteria.
~ d. Plug Welds The TRT investigated alleged generic problems regarding uncontrolled repairs to holes existing.in pipe supports, cable tray supports and base
, plates in Units 1 and 2. These holes, which had been misdrilled during fabrication, were repaired by plug welds. Since these support's are Seismic l' i K-177
incident had not been performed. The TRT review of Gibbs & Hill Specifica-tion No. 2323-MS-100 indicated that there were inadequate requirements and construction practices for the support of the main steam line during flushing, and for temporary supports for piping and equipment in general. In particular, evaluations to assure the adequacy of temporary supports during flushing and installation were not required. The deficiencies in the analyses, specifications and construction practice identified above constitute a violation of Criterion V of Appendix B to 10 CFR 50. Accordingly, TUEC shall: (1) Modify Gibbs & Hill Specification No. 2323-MS-100, and institute pro-cedures for support of the main steam line during flushing and for temporary supports for piping and equipment in general to assure that the quality of piping and equipment are not affected. + (2) Perform an assessment of stresses in the portions of the Unit 1 - loop 1, main steam and feedwater lines that were affected in the sequence of events involved during their initial installation, 4 flushing and final installation. Conditions requiring stress analysis are: (a) Flushing condition when the lines were full of water and temporary supports had sagged or settled. (b) Disconnecting condition when vibrations of the temporary line could have occurred. , (c) Lifting condition when forces were applied by the polar crane and come-alongs. These assessments shall be based on appropriate piping configurations involved. ' (3) Perform a non-destructive examination of locations in the Unit 1 . b loop 1, main steam and feedwater piping where stresses were exceeded during the conditions of concern in a. through c. above. l (4) Review the existing baseline UT examinations for those portions of ' ! the Unit 1, loop 1, main steam and feedwater involved in all the conditions of concern in a. through c., above, for unacceptable inoications. (5) Review records of hydrostatic testing of the main steam and feedwrter line to verify the quality of piping involved in the incident. (6) Provide similar assessments for circumstances involved in a lifting incident identified during the TRT inspection for the Unit 1, loop 4, main steam line. . i K-179 .
j f 1 I (4) Finally, verify during. future Unit I hot functional testing that
- completed modifications to the RPVRI support ring now alicw j adequate cooling air flow.
I The TRT noted that control of debris in critical spaces between . components and/or structures was identified as an issue, both in the - investigation of this allegation and the civil / structural area item II.c (Maintenance of Aire Gap Between Concrete Structures), contained in Darrell'G. Eisenhut's September 18, 1984, letter to TUEC. Accord-ingly, TUEC shall also: ' (1). Identify areas in the plant _ having critical spacing between components and/or structures that are necessary for proper func- > tioning of safety-related components, systems or structures in which unwanted debris may collect and be undetected or be dif-ficult to remove; (2) Prior to fuel load, inspect the areas and spaces identified and ' remove debris; and, (3) Subsequent to fuel load, institute a program to minimize the collection of debris in critical spaces and periodically inspect th.ese spaces and remove any debris which may be present.
- b. Polar Crane Shimming i The TRT investigated the installation of the polar crane rail support i system by visual inspection, review of associated documentation, and '
discussions with TUEC representatives and their contractors. Region IV Inspection Report 50-445/84-08; 50-446/84-04 and Notice of Violation,
- dated July 26, 1984, documented that gaps on the Unit 1 polar crane l bracket and seismic connections exceeded design requirements. In -
L Texas Utilities Generating Company responses of August 23, 1984, and September 7, 1984, the gaps were attributed to crane and bolting . i self-adjustment resulting from crane operation. A site design change (DCA-9872. Revision 4, dated August 24,1984) was issued to document the acceptability of the gaps in excess of 1/16 inch which were identified in the above NRC inspection report, j During further investigation of the allegation that shtms for the
- rail-supoort system of the polar crane had been altered during i
installation, the TRT observed gaps which may have been excessive , between the crane girder and the girder support bracket. Detailed specifications addressing the gap tolerance in the girder seat con-nections did not exist; however, Gibbs & Hill letter GHF-2207, dated 1 November 28, 1977, stated that the " seated connections will not require shimming since the area in bearing is at least the width of the bottom flange of the crane girder." Contrary to this Gibbs & Hill assumption,.the TRT observed nine girders with gaps which K-181 ;
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2 YlTLE AND SstT.TLE o atCartENT 5 ACCissiom Nuwstm Safety Evaluation Report related to the operation of Ccnd.; "cc.t t03m Flectric Station, Units 1 and 2
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$ AUTMQMt3' T. DATE REFORT IS8ut0 WONTM vtAA FEBRUARY 1985
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- 13. $UPPL&wthfA4Y 43Til Docket Nos. 50-445 and 50-446 1<4 AS$ TRACT d209 we,WB er mJ#
Supplement No. 8 to the Safety Evaluation Renort for the Texas Utilities neneratinq Company application for a license to onerate the Comanche Deak Stean Electric Station located in Somervell County, Texas has been .jointiv nrenared by the M* ice of Nuclear Reactor Regulation and the Technical Review Team of the U. S. Nuclear Dequlatorv Commission. This Supplement provides the results of the staff's evaluation and resolution of annroximateh 30 technical concerns and allegations relating to civil /strnctural and miscellaneous issues regarding construction and olant readiness testing oractices at the Ccnanche Peak facilitv. 15e Etv vv0aL3 ANO occuwtNT ANALvses ito Ot$cMirrges 16 &va8LA46biTV ST Ait ut NT 17 StCums t e C LA5&#t sCA TICN 9 Nuwstf407PAGES U ATflFIED Unlimited 16 SE cumef v C& A15df iCATiON 20 Pa'C4 UHt1Tfs t fied 3 +
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