ML20148H173
| ML20148H173 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 01/21/1988 |
| From: | Charemagne Grimes NRC OFFICE OF SPECIAL PROJECTS |
| To: | Counsil W TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| References | |
| NUDOCS 8801270149 | |
| Download: ML20148H173 (40) | |
Text
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[gML Ff Cgko UNITED STATES s
y
'g NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555 j
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v January 21, 1988 Docket Nos. 50-445 50-446 l
Mr. William G. Counsil Executive Vice President TV Electric Skyway Tcwer 400 horth Olive Street, L.B. 81 Dallas, TX 75201 l
Dear Mr. Counsil:
SUBJECT:
REVIEW 0F TECHNICAL CONCERNS AND ALLEGATIONS The Comanche Peak Safety Evaluation Report (SER), NUREG-0797, Supplement No. 7 published in January 1985, Supplement No. 8 published in February 1985, and Supplement No. 10 published in April 1985 dealt with technical concerns and allegations in the areas of electrical / instrumentation, civil / structural, and mechanical / piping, respectively.
The safety evaluations presented in these SER supplements were performed by review groups of NRC's Technical Review Team (TRT). A number of concerns and allecations received by the TP7 were not adcressed in these SER supplements either because they were received late or the technical evaluation was not complete.
These remaining allegations were to be dealt with in future staff evaluations.
The TRT technical evaluation of the remaining allegations in the electrical /
instrumentation ard civil / structural technical areas resulted in the conclusion that there are open issues that remain to be resolved.
Enclosed are the staff's technical evaluations for ten groups of allegations (2 electrical and 8 civil /
structural) where open issues remain.
To provide you with a more comprehensive background to aid in your addressing each open issue, the enclosed evaluations contain the staff assessment of all the allegations within each group although the open issue may not apply to every allegation within a group.
The enclosed technical evaluations were completed by the TRT reviewers who were involved in the evaluations documented in earlier SERs prior to the formation of the Office of Special Projects (OSP).
Issuance of these evaluations was delayed due, in part, to internal NRC staffing reassignments.
However, we understand that, for the most part, the open issues associated with the allega-tions have been brought to the attention of TV Electric. OSP staff has reviewed the enclosed evaluations to assure that they are consistent with ongoing staff' reviews, inspections, and audits.
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050004 5 PD
i TV Electric
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We request that you respond to all the open issues identified in the enclosed technical evaluations within 60 days of the receipt of this letter.
These open issues are identified separately in Paragraph 4 for each group of allegations.
The staff reccenizes that most of the cpen issues are covered by ongoing ana/or completed TV Electric programs. Where this is the case, ycur response should clearly reference these programs ano describe how they address the issue. Where the issue is not addressed by an onooing program, please provide to us your plans and schedule for adcressing the issue.
Should you have any questions or need further clarification, please contact ne (301-402-3299) or Phillip F. McKee (301-492-3301).
Sincerely, Christoph(er I. Grimes, Director Comanche Peak Project Divisien Office of Special Projects
Enclosure:
Technical Evaluation cc: See next page l
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4 W. G. Ccunsil Comanche Peak Steam Electric Station Texas Utilities Electric Ccapany Units 1 and 2 cc:
Jack R. Newman, Esq.
Asst. Director for Inspec. Prcgrams Newman & Holtzinger, P.C.
Comanche Peak Project Division Suite 1000 U.S. Nuclear Regulatory Commission 1615 L Street, N.W.
P. O, box 1029 hashington, D.C. 20036 Granbury, Texas 76048 Robert A. Wooldridge, Esq.
Regional Aaministrator, Region IV Worsham, Forsythe, Sampels &
U.S. Nuclear Regulatory Consission Wooldridge 611 Ryan Plaza Drive, Suite 1000 2001 Bryan Tower, Suite 2500 Arlington, Texas 7601.
Dallas, Texas 75201 i
Lanny A. Sinkin Mr. Homer C. Schmidt Christic Institute Director of Nuclear Services 1324 North Capitol Street Texas Utilities Electric Company Washington, D.C.
20002 Skyway Tower 400 North Olive Street, L.B. 81 Ms. Billie Pirner Garde, Esc.
Oallas, Texas 75201 Government Accountability Project Midwest Office Mr. Robert E. Ballard, Jr.
104 Eest Wisconsin Avenue Director of Projects Appleton, Wisconsin 54911 Gibbs and Hill, Inc.
11 Penn Plaza New York, New York 10001 David R. Pigott, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street Mr. R. S. Howard San Francisco, California 94111 Westinghouse Electric Corporation P. O. Box 355 Anthony Z. Roisman, Esq.
Pittsburgh, Pennsylvania 15230 Suite 600 1401 New York Avenue, NW Renea Hicks, Esq.
Washington, D.C. 20005 Assistant Attorney General Environmental Protection Division Robert Jablon P. O. Box 12548, Capitol Station Bonnie S. Blair Austin, Texas 78711 Spiegel & McDiarmid 1350 New York Avenue, NW Mrs. Juanita Ellis, President Washington, D.C. 20005-4798 Citizens Association for Sound Energy 1426 South Polk George A. Parker, Chairman Dallas, Texas 75224 Public Utility Committee Senior Citizens Alliance Of Ms. hancy H. Williams Tarrant County, Inc.
CYGNA Energy Services 6048 Wonder Drive 2121 N. California Blvd., Suite 390 Fort Worth, Texas 76133 Walnut Creek, CA 94596
0 i
o W. G. Counsil Comancho Peak Electric Station Texas Utilities Electric Company Units 1 ard 2 cc:
Joseph F. Fulbright Jack P. Newman, Esq.
Fulbright & Jaworski Newman & Holtzinger, P.C.
1301 McKinney Street Suite 1000 Houston, Texas 77010 1615 L Street, N. W.
Washington, D.C. 20036 Roger D. Walker fianager, Nuclear Licensing Texas Utilities Electric Company Skyway Tower 400 North Olive Street, L.B. 81 Dallas, Texas 75201 Mr. Jack Redding c/o Bethesda Licensing Texas Utilities Electric Company 3 Metro Center, Suite 610 Eethesda, Maryland 20814 William A. Burchette, Esq.
Counsel for Tex-La Electric Ccoperative of Texas Heron, Burchette, Ruckert & Rothwell Suite 700 1025 Thomas Jefferson Street, NW Washington, D.C.
20007 James P. McGaughy, Jr.
6DS Associates, Inc.
)
Suite 720 j
1850 Park. tray Place i
Marietta, Georgia 30067 Administrative Judge Peter Bloch U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Elizabeth 8. Johnson Administrative Judge l
Oak Ridge National Laboratory P. O. Box X, Building 3500 Oak Ridge, Tennessee 37830 Dr. Kenneth A. McCollom 1107 West Knapp Stillwater, Oklahoma 74075 Dr. Walter H. Jordan c/o Carib Terrace Motel 522 N. Ocean Boulevard
' Pompano Beach, Flor.ida 33062
ENCLOSURE TECHNICAL EVALUATION COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AhD 2 TECHNICAL CONCERNS AND ALLEGATIONS I.
Electrical and Instrumentation A.
Electrical Cable Terminations (Allegation Numbers: AE-56 and AE-57) 1.
Characterization:
It was alleged that:
a.
The use of smaller gage wire in the fire alarm panels nr.y overicad the wire capacity and cause a fire (AE-56).
b.
A design change authorization (DCA) that authorized cable splicino in a control panel was issued contrary to procedure (AE-57).
2.
Assessment of Safety Sionificance:
AE-56 Texas Utilities Electric Company (TV Electric) Specification 2323-ES-17B for fire alarm panels originally required that all panel wires be ho. 14 American Wire Gage (AWG) minimum.
These fire alarm panels are nianufactured to rigio flammability restric-tion criteria in accordance with National Electric Code (NEC) and Underwriter Laboratories (UL) requirements.
The vendor requested a deviation from the No. 14 AWG minimum panel wiring requirement on the grounds that No.14 AWG wire wculd not fit into the connector.
The vendor proposed the use of No. 14 AWG minimum wire for the 120-volt alternating current (VAC) power wiring, No. 18 AWG minimum for the internal power bus wiring, and No. 22 minimum for printed circuit board (PCF2) interconnections.
The hRC Technical Review Team (TRT) cetermined that, on the basis of ampacity alone, the vendor's proposed use of the No. 14 AWG, No. 18 AWG, and No. 22 AWG wiring was fully justified.
Ko. 22 AWG minimum wires, which were proposed for use in the PCB inter-connections, have a rating of 2 amperes (A); however, the vendor states that they normally carry a current load of 70 milliamperes (mA).
lhe vendor also proposed using No. 18 AWG minimum wires, which have a rating of 6A but normally carry a current load of 0.5A, for the internal power wiring to the lamp test bus.
The No. 14 AWG wiring, proposed for internal power busing and dry cutput contacts, is rated ISA and is protected by 15A circuit breakers.
Gibbs & Hill (G&H) initiated a revision to Specification 2323-ES-170 incorporating the changes ~ tor the vendor-furnished equiptrent wiring; however, the revision process was not completed.
TV Electric issued Design / Engineering Change /Deviatinn Request (DECD) N-661, Rev.1, to incorporate the required changes into Specification 2323-ES-178.
Revision 1 to DECD N-661 was issued twice.
The first issuance (July 14, 1980) included changes to allow the use of tefzel-type wire and both the 18 AWG n.inimum and 22 AWG minimum requirements.
However, the second tirre DECD N-661, Rev.1, was issued (October 20, 1980), the 22 AWG minimum requirenient was not included. TU Electric j
stated that this requirement had not been included because they did not deem it fit to include criteria affecting the vendor's low voltage wiring in a 1U Electric specification.
The IRT dia not acree with the TV Electric decision not to include the 22 AkG minimurn requirement in DECD N-661, Rev.1.
Criterion VII of Appendix B to 10 CFR 50 states that the applicant is required to establish measures
"...to assure that purchaseo material, equipnient, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents." Any excep-tions taken by the vendor must be carefully evaluated, and the applicant must either deny them or incorporate them into the specification.
AE-57 Allegations AE-13, AE-18, and AE-22 in Safety Evaluation Report Supplement (SSER) No. 7 addressed concerns related to butt splices I
in panels.
In a follow-up interview with the NRC staff, the allecer stated that DCA 19264 authorized the splicing of 55 multi-conductor control cables in panel No. 1-CR-14.
This panel is located at the 830-foot elevation in the Main Control Room.
The alleger was concerned that DCA 19264 was indicative of widespread cable splicino at Comanche Feak Steam Electric Station (CPSES).
Butt splicing cables in panels is permitted on a limited basis, as specified in Section B.1.5.2.4 of Amendment d4 to the CPSES Final Safety Analysis Report (FSAR).
The NRC staff reviewed TUEC's justification for permitting butt splices inside panels ar.d deter-mined that the practice was acceptable on a limited basis, subject to the ccnditions stated on page J-28 of SSEf! Mo. 7.
The TRT reviewed the history of the events that led to the subject DCA and noted that only 6 of the 55 multi-conductor cables were safety-related.
The TRT also determined that DCA 19264 was written to respond to NRC-required human factors consideratiens due to the physical arrangement of components on the panel.
The TRT inspected panel No.1-CR-14 to verify the splices on 55 cables and found that the splices were not adequately stacgered, which violated one of the conditions stated on page J-28 of SSER No. 7.
j j
3.
Conclusions and Staff Positions: With respect to Allegation AE-56, the TRT concluoed that adequate reconciliation is lacking between TV Electric's Specification 2323-ES-178 and the vendor's information, j
In addition, the TRT concluded that Allegation AE-57 raises concerns similar to those stated in Item 5 (first tullet) on page J-30 of l
SSER No. 7.
4.
Open Issue _s_: TV Electric should identify actions taken or needed to be taken to:
a.
Reconcile differences between Specification 2323-ES-178 and the verdor's information.
b.
Assure that the concerns raised by Allegation AE-57 are addressed by actions taken in response to the concerns stated in Item 5 (first bullet) on page J-30 of SSER-7.
Reference Documents
- 1. CPSES Units 1 and 2, Program Pian and Issue-Specific Action Plans, TV Electric Item Nos. I.a.2 and I.a.3, dated October 8,1984.
- 2. Letter from D. N. Chapman (TUGC0) to N. Keddis (Gibbs & Hill) recarding Texas Utilities Gererating Company (TUGCO) Quality Assurance (QA} Audit Notification, QA Audit File TGH-14, dated May 6,1980.
- 3. Letter from D. N. Chapman (TUGCO) to N. Xeddis (Gibbs & Hill) regarding the results of TUGC0 OA Audit Report No. TGH-14, dated June 24, 1980.
- 4. Letter from N. Keddis (Gibbs & Hill) to D. N. Chapnan (TUGCO) regarding corrective action response to TUGC0 QA Audit Report No. TGH-14, dated July 18, 1980.
S. Letter from N. Kedcis (Gibbs & Hill) to P. N. Chopman (TUGCO) regarding supplemental information on TUGC0 QA Audit Report No. TGH-14, dated August
- 21. 1980.
- 6. Letter from D. N. Chapman (TUGCO) to N. Keddis (Gibbs & Hill) regarding closing of all items addressed in TGH-14,
- 7. TRT interview with J. Ellis ar.c M. Walsh, pages 81-82, dated November 7, 1984.
- 8. Letter frcm P. E. Valentino (Alison Control, Inc.) to S. Martinovich (Gibbs
& Hill) regarding Alison standards for Wiring, dated March 3, 1985.
- 9. Letter from R. E. Ballard, Jr. (Gibbs & Hill, to J. B. George (TUGCO),
regarding Gibbs and Hill's review and conclusion on the use of wire size smaller than No. 14 AWG in fire detection panels furnished by Alison Controls, dated March 19, 1985.
- 10. Design / Engineering Change / Deviation (DECD) Request No. N-611, Rev. 1.
- 11. Letter f rcm P. E. Valentino ( Alison Control, Inc.) to H. Rock (Gibbs &
Hill) requesting a deviation for the use of No. 22 AWG wire in fire detection systems, dated June 17, 1980.
- 12. Specification 2323-ES-178, Rev. 1, dated June 1, 1979.
- 13. DCA No. 7211, Rev. 4, dated September 19, 1983, revising Specification 2323-ES-178 per DECD N-611, Rev. 1,
- 14. Field Design Change & Review Status Log, "Affected Document Update Report, regaroing Specification 2323-ES-178," dated September 12, 1985,
- 15. Vemorandum, M. Srinivasan (NRC) to B. Youngblood (NRC), dateo July 30, 1985.
- 16. Safety Evaluation Report, NUREG-0747, Supplement No. 7. January 1986.
- 17. Final Safety Analysis Report for the Comanche Peak Steam Electric Station.
s B.
Electrical honconfomance Report Activities (Allecation Numbers: AQE-34, AQE-35, AQE-37 and AE-55)
Characterization:
It was alleged that the validity of the 1.
tion ano aisposition of electrical nonconformance reports (genera-NCRs) was suspect. These allegetions pertained to varicus concerns involving the NCR program and include:
a.
Prevalent"use-as-is"dispositionsofNCRs(AQE-34andAQE-35).
b.
Traceability of "Q" items (non-Q fuse blocks were installed i
where "Q" blocks were required) (AQE-35).
l c.
Disposition of hCR on termira.1 block reworks was cuestionable (AQE-37).
Four safety-related Class 1E meters (two emergenc cenerator (ECG) wattmeters and two EDG varmeters)y diesel d.
were removed from the main control beard, modified by an unapproved facility, then reinstalled in the main control board without procedures (AE-55).
2.
Assessment of Safety Sicnificance:
AQE-34 leie NRC Technical Review Team (TRT) reviewed information from the alleger concerning craft personnel usinc an all-thread rod to disicdge a Bisco seal.
The alleger stated that the rod was bent while the seal was being dislodged and that the tape used to cover threads at one end was dangling from the rod when it was retrieved.
NCR No. E84-00673 was initiated and was cispositioned use-as-is" following a field inspection.
The alleger'also stated that there were no adequate procedures for the retreval of penetration seals.
The TRT found that the primary purpose for removing the Bisco seals at the CPSES was to obtain cable slack. The predominant scal material has a foam-like texture and is not difficult to dislodge.
Electri-cians usually puncture the seal material with a wooden dowel, then use a bent copper rod to break it off.
Both the dowel and the rod are taped over with several layers of black electrician's tape.
The TRT reviewed the pertinent documents ard determined that dis-looging and retroving a Bisco seal with an all-thread rod that had been properly tapso at one end would not, by itself, darrage the cables. Although the rod could easily be bent during this process, a bent rod does not trean that the cables would be damaged. The TRT found that the all-thread rod was taped to minimize damage to the cables and determined that tape dangling from the rod was not in-dicative that the threads on the rod were uncovered.
1 AQE-35 The TRT reviewea information from the alleger concerning the in-stallation of non-Q fuse blocks in lieu of "Q" fuse blocks.
The i
TRT traced an instance where non-Q fuse blocks were used in the testing phase of equipment installation and startup. The TRT found that the non-Q materials were subsequently replaced with the "Q" fuse blocks and that all work was systematically executed and pro-perly documented.
The TRT also reviewed the "use-as-is" disposition cf the NCR associ-ated with Buchanan terminal blocks.
The allecer stated that these i
terminal blocks were not properly called out per their "Q" materials procurement documents when they were used as replacement parts in nonsafety-related equipment applications.
The TRT examined the NCR and agreed with the "use-as-is" disposition.
The terminal blocks are, in effect, safety-related materials pur-chased in bulk quantity for use in safety-related equipment. One of these terminal blocks can be used to replace any similar Buchanan terminal block in the plant, and using then in nonsafety-related equipment does not invalidate their use as "Q" material.
_A_Q E - 3 7 The TRT reviewed information from the alleger regarding terminal block rework and examined No. NCR E83-03239.
The alleger statea that contrary to procedure, terminal block rework was completed under Startup Procedure CP-SAP-6, instead of under Construction Procedure CMP 6.10, Rev. 10.
The TRT four.d that rework on the terminal block was covered under Startup Work Authorization (SWA) l 14163, which ensured retesting as per paragraph 4.1.1(2) of CP-SAP-6.
The TRT also determined that wbsequent testing was per-fermed. A review of pertinent documents showed that the rework was a startup activity, not a construction activity; therefore, the NCR was properly dispositioned.
AE-55 The TRT noted that this allegation was documented in NCR No.
E81-00088 on March 30, 1981.
This NCR stated that fcur safety-related Class 1E meters (two EDG wattmeters and two EDG varv.eters) were removed from the Unit 1 Main Control Board without the use of procedures, that they were sent offsite to an unapproved facility for trodification, and they they were reinstalled in the control board without the use of procedures.
The NCR also stated that, as a result of this work, the quality status of these four instruments and the control board was indeterminate.
A review of the documentation related to the EDG varmeters and wattmeters and interviews with on-site personnel revealed that the disposition of the original NCR directed that the meters be replaced with r.ew meters in accordance with G&H Specification 2323-MS-605, Rev. 1.
This would have resolved the questionable status of the meters; however, it was later discovered that General Electric Con.pany (GE), the meter manufacturer, no longer supplied these meters with Class Ir. certification.
Consequently, on August 10, 1982, the disposit'on of the NCR was revised.
)
s Revision 1 to NCR No. E81-C0088 stated that GE would no longer provic'ci Class 1E meters; therefore, no replacements were available.
The revision also stated that, since an unapproved facility had i
made a simple scale change, the TV Electric calibration facility was l
to inspect and calibrate these rneters to assure their cperability and integrity.
The revision made no tr.ention of recertifying cr deleting tne Class 1E quality of the meters.
On June 4 and June 6,1984, while test'ag the Unit 1 FOGS, the varmeters for bcth EDGs incorrectly indicated hiah. Test Deficiency Reports (TDRs) 2825 and 2835 documented the p.oL7ert The resolution to this problem was to use the vanneters from the Unit 2 control board in Unit 1.
The Unit 2 varr:0ters h:J been modified by an authorized vendor (Reliance Electric Corrpany) in accordance with DCA 9714, Rev. 1.
The TRT reviewed the docurrentation und found no problem with either the Unit 2 varmeters cr the Unit 2 wattmeters.
i Althcugh the Unit 1 control board had been supplied with two pro-perly certified, operable Class 1E vartneters, the original problem with the quality of the two wattnoters, which had been listed in NCR ho E81-00088, Rev. 1, was not addressed.
The TRT reviewed G&H Specification 2323-MS-605, Rev. 2, and noted that the meters in questiun were still designated as Class 1E.
The TRT questioned TV Electric as to how the Class 1E certification on the Unit 1 wattmeters had been re-established.
TU Electric responded that vendor recertification was not required and that future replacement treters will be handled as commercial quality, as described in Procedure TN-PR-3, Rev. 2. "Classificatior: of Safety-Related Replacement Parts." This procedure defir.es cormercial quality 1
(QA Ccde C) as a level of quality assigned to parts that have been detennined to contribute to the safety function of safety-related equiptrent (such as the EDGs) and that are of simple construction.
made to national standards, widely used in the industry, and trass-produced by an industry ccmpetitive enough to force quality through commercial pressures.
\\
The TRT reviewed TNE-Po-3 and found that the procedure provided for the classification of safety-related parts as a function of the spare and replacement part procurerrent process.
However, the pro-cedure did not address reclassification of equiptent that had already been procured and/or installed.
TNE-DC-6, Rev.1, "Equipment Classification Proceoure," appeared to the TRT to provide instructions for reclassifying components such as wattmeters.
However, when the TRT requesteo occumented evidence to show an engineering analysis and a formal downgrading cf the meters te non-Class IE, none was provided.
G&H Specification 2323-MS-605, Rev. 2, continued to identify the meters as being Class 1E, which indicated that design control treasures had not been implemented.
On May 10, 1985, meters W-1EG1 anc W-1EG2 were designated Cless 1E, as specified in G8H Specification 2323-MS-605, Rev. 2.
A representative from TV Electric stated that:
"No vendor recertification is re-quired.
Engineering dispositioned the NCR based on the fact that GE will not provide Class 1E meters.
Future replacements will be crdered as commercial cuality." Nc documentation was made avail-able to the TRT to support downgrading the meters from Class IE to non-Class 1E (ccomercial) status.
3.
Conclusions and Staff Positions:
AQE-34 Based on an inspection of cables in penetrations after the Bisco seals were dislodged, the TRT concluded that o determination could be n,ade regarding the adequacy of the cables inside the penetrations during and after the Bisco seal removal operation.
The TRT reviewed the "use-as-is" disposition of NCR E84-00673 and determined that the integrity of the cables could be determined; thus, the hCR was pro-perly dispositioned.
The TRT also determined that adequate pro-cedures existed for the removal of penetration seals.
Accordingly, the TRT concluded that this allegation was not sub-stantiated and had neither safety significance nor generic in:pli-cations.
AQE-35 The TRT reviewed additional input from the alleger regarding non-Q fuse blocks that were allegedly installed in lieu of "Q" fuse blocks and determined that all work was systematically executed and pro-perly documented.
Based on its review of the NCR associated with Buchanan terminal blocks, the TRT agreed with the "use-as-is" dis-position.
The IRT concluded that there were no improper dispositions of the NCR and that the allegation was not substantiated.
This agreed with a similar conclusien in SSER No. 7 that AQE-35 could nut be substantiated.
Accordingly, the TRT concluded that this allegation had neither safety significance nor generic implications.
AQE-37 Based on the review of additional information from the alleger and dn examination of NCR No. E83-03239, the TRT determined that rework on the terminal block was covered under SWA 14163, which ensured retest of the terminal block.
The TRT also determined that rework is a startup activity, not a construction activity.
Therefore, the TRT concluded that this allegation eculd not be substantiated because the NCR was properly dispositioneo and there was no viola-tion of procedure.
Accordingly, the TRT concluded that this allegation had neither safety significance nor generic implications...
1 AE-55 The TRT determined that NCR No. E81-00088 documented removal of four meters and that TU Electric's quality assurance program 4
adequately identified and documented the problem. However, the engineering disposition of NCR E81-0088 was inadequate since design j
controls in accordance with G8H Specificatien 2323-MS-605, Rev. 2, were not properly implemented to preserve Class 1E certification of the meters or to formally acwngrade thene.
The TRT concluded that, since the varmeters for Unit I were eventually
)
repaired and recertified as Class IE by an approved facility, only the eouipment qualification status of watteeters was questionablo.
4.
Open Issues: TU Electric should identify actions taken or needed to be taken to:
a.
Provide adequate design controls to properly certify the EDG wattreters and varmeters.
b.
Establish the equipment qualification status of the wattmeters for Unit 1.
Reference Documents:
- 1. IlCR No. E84-00673, dated March 5, 1984,
- 2. Inspection Peport No. EC-10029591, dated March 5, 1984.
- 3. Testimony of alleger, Granbury, Texas, October 30, 1984, pages 15-30.
- 4. Brown & Root, Inc. (B&R) Procedure No. Cl-CPM 6.12, Rev. 3, "Penetration Seals Area Release," dated June 8, 1983.
- 5. B&R Procedure No. Cl-CPM-6.12, Rev. 4, "Area Release for Penetration Seals and Removal of the Seismic Gap Flashing," dated Cctober 24, 1984.
- 6. Bisco Procedure No SP-107, Rev. 2 "Repair ano hawork of Bisco Silicone Base Penetration Seals," dated June 2, 1982.
- 7. NCR No. E83-01218A, dated August 24, 1983.
- 8. Construction Operation Travelers:
No. EE83-0214-7404, dated February 14, 1983 No. EE83-0215-7404, dated February 14, 1983 No. EE83-0450-7404, dated May 20, 1983 No. EE83-0451-7404, dated May 20, 1983
- 9. NCR No. E83-02388, Rev. 1, dated September 9, 1983.
- 10. Testimony of Alleger, Granbury, Texas, October 30, 1984, pages 35-39.
- 11. TUSI office memo from P. B. Stevens to I. Vogelsand, "Purchase and Installation Requirements for Fuses," dated September 22, 1983.
- 12. NCR No. E83-03239, dated December 13, 1983,
- 13. Startup Administrative Procedure, CP-SAP-6, Rev. 9, "Control of Work on Station Components After Release from Construction to TUGCO,"
dated October 24, 1983.
- 14. TNE-PR-3, Rev. 2 "Classification of Safety-Related Replacement Parts."
- 15. TNE-0C-6, Pevision 1, "Equipment Classification Procedure "
- 16. NCR No. E81-00088, dated March 30, 1981.
- 17. NCR No. E81-00888, Rev. 1, dated August 10, 1982.
- 18. Letter to W. F. Smith (NRC) frcm S. M. Franks (TV Electric), "Response to Questions Pertaining to Disposition of NCR E81-00088, Rev. 1 (with attachments), dated March 27, 1985.
i
- 19. Gibbs & Hill Specification 2323-MS-605, Rev. 1, DECD S-1700 (also Revision 2).
i
- 20. DCA 9714, Rev.1, "Change Diesel Generator Vanneter nr.d Wattmeter Scales."
- 21. "Sumary of Documents Relatino to Diesel Generator Wattmeters and Variteters."
- 22. TDR 2825, dated June 4, 1984.
- 23. TCR 2835, dated June 6, 1984.
- 24. Attachment B uf CP-SAP-3, Rev. 5, "System / Subsystem Turnover," dated I
July 3, 1980.
- 25. Startup Werk Authori:ation 14163
- 26. Safety Evaluation Report, NUREC-0797, Supplement No. 11, May 1985.
l l j
II. Civil and Structural (C/S)
A.
Reinforcing Steel (Alleaation Nos: AC-56, AC-57, AC-58, AC-59, AC-60 and AC-61) 1.
Characterization: The Citizens Association for Sound Energy (CASE) questioned the adequacy of the disposition of the following non-conformance reports (NCRs):
a.
NCR No. C-82-00523, reinforcing steel missing from the Unit I containment structure wall (AC-56).
b.
NCR No. C-811, 46 No. 9 reinforcing bars.were-omitted from a wall in the Excess Letdown Heat Exchanger Roca in the Unit 1 containment structure (AC-57).
c.
NCR No. C-1314, 57 No. 5 dowels were bent and 10 No. 5 dowels were broken off at the concrete in the Unit I containment structure (AC-58),
in addition, CASE raised questluns concerning the fact that:
d.
no analyses were performed justifying the omission of reinforcing steel throughout the plant (AC-59), and e.
these omissions of reinforcing steel are another example of QA/QC failure ( AC-60).
- Finally, t.
CASE raised a concern questioning the conclusion drawn by a Recion I" investigation of an allegation concerning the omission of horizontal "tie" reinforcement in the Unit I containment structure wall (40-61).
i There concerns were raised in a letter from Juanita Ellis, i
PresioentofCASE,toVincentNoonan(NRC)catedDecember1, 1984 The NRC Technical Review Team (TRT) did not meet with Mrs. Ellis to discuss these concerns after receiving this letter, since the TRT believed the issues to be sufficiently clear to allcw its investigation to proceed.
2.
Assessment of Safety Sionificance:
AC-56 On May 3, 1982, Brown & Root (B&R) issued NCR No. C-82-00523, l
documenting the omission of No. 6 "Z"-shaped shear ties from a concrete placement within the construction access opening in the Unit I containment structure wall.
The concrete of Placement No. 101-5805-036 had been placed to elevation 823 feet, 6 inches without the shear bars having been placed at the 820-foot, 8-inch cnd 822-foot, 6-inch elevations.
Gibbs & Hill (GaH) was informed of this omission on May 4, 1982, by B&R engineering.
GSH en-gineering perforraad an analysis (calculation book SRB-1220, Set 6) and concluded that the as-constructed cordition was acceptable,.
provided that one additional row of No. 6 shear ties be placed at the 823-foot, 6-1/2-inch elevation to compensate for the missing two rows of the "Z"-shaped shear ties.
B&R engineering then issued Pesign Change Authorization (DCA) No.13533 on May 17,1982, to add the additional row of No. 6 shear ties. On May 18, 1982, as per the disposition of NCR No. C-82-00523, the placement of the additional reinforcirg steel was inspected ard accepted.
The TRT reviewed the analysis performed by G&H.
To ensure that the ' analysis was correctly performed and to resolve questions raised during its review, the TRT audited these calculations and all supporting documentation at the G8H office in New York, New York on May 23, 1986.
The TRT questions were resolved at the audit, and it was ccncluded that the actions taken to compensate for the missing shear ties were acceptable.
AC-57 A concern about the omission of 46 No. 9 reinforcing dowels in a j
wall between the Excess Letdown Heat Exchanger Room and Steam Gen-erator Compartment No.1 in Reactor Building No. I was previously addressed in SSER No. 8, Civil / Structural (C/S) Category 6, page K-49.
The disposition of NCR No. C-811, dated October 31, 1977, directed that the 46 nissing reinforcing dowels be placed by drilling holes 2-1/2 inches in diameter and 48-1/2 inches deep, placing the rebar, and filling the remaining cavity with grout.
The TRT based its earlier assessment on the assumption that all 46 reinforcing dowels were drilled and grouted in place.
Subsequent to the issuance of S$ER No. 8, the TRT obtaineo Design Chance / Design Ceviation Authori-ration (DC/DDA) ho. 696, which stated thet, due to rebar congestion and embedments, it was not possible to drill all of the required 46 holes or to obtain the specified embedded depth of 48-1/2 inches on some of the drilled holes.
B&R construction was able to drill 35 holes, of which only 3 had a depth greater then 48 inches.
l Occause the specified hole depth could riot be met, R. W. Hunt Co.,
which operated the cLocrete testing laboratory on site, performed pullout tests on scmples cf grouted rebar to determine the embedment (hole depth) that would provide sufficient anchorcge to develop the strength of a No. 9 reinforcing bar using a i
specified type of grout.
These tests showea that an embedment of 18 inches would be sufficient. Of the 35 holes orillec, 5 had a depth of less than 18 inches; these holes were not used 4
and were filled with grout.
Accordingly, out of a total of 46 missing reinforcing dowels, 30 were placed.
The DC/CDA further stated that calculations were made confirming the adequacy of the reduced quantity of reir. forcing steel to perform the required design function.
The calculations were contained in calculation book No. SRB-1200.
The TRT learned that lu Electric had requested these calculations from G&H on December 26, 1984, as part of their effort to research all cases of reinforcing steel omission for proper engineering dis-position.
This action stemmed from the TRT's request for more in-forroation pertaining to the issue of missing rebar in the reactor a
cavity wall in Unit 1, which was identifiea in C/S Category 6.
TU Electric was informed by G&H that the relevant calculations in book SRB-120C could not be traced and that new calculations were beino prepared.
The TRT reviewed the new calculations which shewed that the wall would acequately carry the loads considered. Since a steam cenerator lower lateral restraint beam frames into the wall in this 5rea, and because a question was raised concerning the adecuacy of the steam generator lateral restraint beams (C/S Category 27), the TRT inquired as to whether these calculations considered loads applied to the wall by thermal growth of the beam. The TRT was in-formed by a G8H representative that this loading was not considered.
As a result of the review associated with Paragraph II.0, the 1RT was aware that a NASTPAN analysis, which included the lower lateral j
restraint beams, had been perfor.ted.
As indicated in Paragraph II.0, TU Electric committed to prcvide an assessment of the overall aesign adequacy of both the upper and lower steam generator lateral restraint beams for all load combinations. One of the components to be reviewed included'the concrete walls into which the beams frame. TU Electric's review of the wall in question bere, should take into account the effect of the missing rebar.
AC-58 NCR No. C-1314 was issued on January 17, 1979.
This NCR reported that, due to the installation and rerr. oval of shoring and scaffolding in the area, 57 No. 5 dowels in Reactor Building
)
Mo. I had been bent, and ar.other 10 No. 5 dowels were broken cff at the concrete.
These dowels were located at the 808-foot elevation between columns 9 and 10 ano were intended to serve as reinforcement for the foundations of the reutron wall cooling units.
The dowels protruded approximately 2 feet from the 808-foot elevation slab, eno were intended to be bent later to meet fctndation requirements.
1 On July 2, 1979, CC/DDA No. 5080 was issued.
DC/DDA No. 5080 detailed the foundation requirements and noted that the existing i
No. 5 dowels should be bent as required.
Following the issuance of the DC/DDA, NCR No. C-1314 was updated instructing B&R construc-I tion to heat and bend any existing (57) No. 5 dowel in accordance with B&R procedure CCP-18 and replace any cracked or broken No. 5 dowel by drilling and grouting per B&R procedure CCP-12.
The TRT reviewed the inspection reports documenting the heating and bending of the bars and the drilling and grouting of the replacement dowels.
These inspecticn reports indicate that the work was performed in a satisfactory manner and according to procedure.
AC-59 in its earlier review of C/S Category 6 and in this present review, the TRT reviewed a total of eight documented cases concerning the omission of reinforcing steel from various concrete placements.
CASE raised a concern that, in many cases, calculations were not performed to justify the omission.
In two of the cases reviewed by the TRT (i.e., the omission of reinforcing steel in the Unit I reactor cavity wall, Item c (1) of C/S Category 6, and in the wall
- 1? -
between the Excess Letdown Heat Exchanger Room and Steam Generator Compartment No.1, Item b,.above), the TRT found that the situation was not resolved in an acceptable manner.
In the rerraining cases, the reinforcing steel was added by drilling and grcuting, and shoring was left in place or calculations were perfonred showing the struc-ture able to carry the design loads as-built.
As mentioned in para-graph b, above, TU Electric is in the process of researching all documented cases of missing reinforcing steel for proper engineerirg disposition. The NRC staff will review TV Electric's re-evaluation.
AC-60 In SSER 8, C/S Category 6, the staff determined that there was an apparent weakness in the qu611ty control program as evidenced by the fact that emitted reinforcing steel was not detected prior to concrete placement.
The TRT C/S Group referred this matter to the TRT QA/QC Group for consideration in their assessment of the overall adecuacy of the QA/QC program.
(See Secticns 2.1 and 2.4 of Appendix P, SSER No. 11.) This apparent weakness of the quality centrol program in the area of rebar placement inspection raises questions concerning how many unidentified cases of missing rebar exist throughout the plant and the implication of this omission en structural integrity.
AC-61 CASE raised a concern questioning the conclusion drawn by a Region IV inspection (Inspection Report 79-25) regarding an alle-gation made in 1979 by a former B&R errployee.
This allegation concerned the omission of horizcntal "tie" reinforcement in the Unit 1 containment structure wall. The Fegion IV inspector con-cluded that the alleger was referrino to a documented case in which horizontal shear reinforcerrent was emitted from a concrete placement near the intersection of the dome and wall in Unit 2.
This conclusion was baseo on the fact that the event occurred shortly before the alleger terminated his errployment; the allega-tion was based on hearsay information relative to events about which the elleger had little or no personal knowledge.
The TRT investigated this allegation (AC-38) in C/S Category 6 and came to the same cenclusion as Region IV.
The TRT based its conclusion on an exanination of the results of the Region IV investigation and on the fact that the rebar placement checklists for the Unit 1 containment structure wall showed all reinforcing steel was placed as required.
Subsequently, the IRT searched a computer printout of all design deficiency reports (DDRs) and NCRs written in the Civil / Structural area from March 1, 1975 to November 7, 1984.
The TRT found two NCRs documenting missing reinforcing steel in the Unit 1 and Unit 2 containment structure walls (NCRs No. C-1653 and NCR ho. C-82-005231 NCR C-82-00523 (addressed in paragraph a, above) was written in 1982, long after
- 1S -
l
the allegation was mcde.
NCR No. C-1653, which documented missing horizontal shear ties in the Unit 2 containment structure wall, is the instance to which the TRT and Region IV believe the alleger is referring.
3.
Conclusion and Staff Position:
AC-56 The TRT concluded that the actions taken to corrcensate for the missing shear ties, as documented in NCR No. C-82-00523, were acceptable and were adequately corroborated by the G&H engineering evaluation.
In addition, the containment passed a structural j
integrity test with no evidence of structural degradation, thus demonstrating its structural safety and adequacy.
AC-57 The TRT concluded that TV Electric has not fully documented the adequacy of the wall between the Excess Letdown Heat Exchanger Room and Steam Generator Compartment 1 in Reactor Building No. 1.
AC-58 The TRT concluded that the action taken to repair the bent and broken reinforcing bars, as docurr.ented in NCR No. C-1314, is acceptable ano that it bes no adverse effect on the structural safety of the foundations.
3.E-E The TRT understood that TU Electric was reviewing all ct the docu-trented cases of rebar omission for proper engineering disposition.
The potential safety significance of this issue cannot be deter-mined until completion of 1U Electric's review.
AC-60 The issue of omitted reinforcing steel was previously noted as a concern in Sections 2.1 and 2.4 of Appendix P, SSER No. 11.
The concern regarding the structural inplications of unidentified cases of rebar emission thrcughout the plant remains to be resolved.
AC-61 The TRT supports the ccnclusion drawn by the Region IV investiga-tien of this issue and concluded that this issue has no structural safety significance.
4.
Open Issues:
TV Electric should identify actions taken or needed to be taken to:
a.
Analyze the as-built condition of the wall betweem Steam Ger,erator Ccmpartment No. I and the excess Letdcwn Heat Exchanger Room in the Unit I containment structure.
b.
Provide an assessment of unidentified rebar omission cases and their implication on the structural safety of the plant.
Reference Documents:
- 1. Drawing 2323-SI-0505,_Rev. 13.
- 2. Drawing 2323-51-0519, Rev. 4.
)
- 3. Drawing 2323-Sl-0520, Rev. 3.
- 4. Drawing 2323-S1-0570, Rev. 4.
- 5. Drawing 2323-51-0521, Rev. 3.
- 6. Drawing SCB-10522, Shts 4, 5.
- 7. Drawing SCB-10519, Shts.1, 5.
- 8. GTN-69753.
- 9. Calculations SMI-102C, Set 2.
- 10. NCR No. C-Sil,
- 11. GHF-2183.
- 12. DC/DDA No. 696.
- 13. TWX-1122.
- 14. NCR No. C-1314.
- 15. OC/DDA No. 5080.
- 16. NCR No. C-82-00523.
- 17. GTN-59187.
- 18. TWX-13520.
- 19. DCA No. 13353,
- 20. Letter from Juanita Ellis (CASE) to Vir. cent Noonrr (NRC), dated December 1, 1984.
- 21. Calculations SRB-122C, Set 2.
- 22. Letter to D. Jeng from P. Bezler "Audit Trip Repcrt" (concerning audit of G&H Calculation Book SRB-122C Set 6 at G8H office! on May 23,1986), dated May 30, 1986.
j
- 23. Safety Evaluation Report, NUREG-0797, Supplement No. 8, February 1985.
- 24. Safety Evaluation Report, NUREG-0797, Supplement ho. 11, May 1985.
- 25. NCR No. C-1653. -..
i B.
Design and Analysis (Allegation Number: AC-69) j 1.
Characterization:
It was alleged that the damage study made to assess the consequences of failure of a nonsafety-Niated system or compcnent un a safety-related system or componenc may not have used appropriate parameters to define the zone of influence (AC-69).
This issue was raised during a meeting between the NRC Technical Review Team (TRT) and the Citizens Association for Sound Energy (CASE) on November 7, 198a.
2.
Assessment of Safety Significarce:
Allegation AC-09 concerns the proper execution of the damage study made to assess the consequences of failure of non-seismic piping and equipment on safety-related systens and components. Of particular concern was the definition of the damage zone, which the alleger describes as a projectile range with a slope of two to Ne (two horizontal to one vertical).
The TRT reviewed the two basic documents which describe the pro-cedures and the criteria for the damage study:
Instruction CP-El-4.0-36, "Control of Seismic and Non-Seismic Ccmponent Inter-action Evaluations," Rev. 3, issued on December 9, 1983; and "Comanche Peak Seismic Interaction Criteria," Rev. 1 dated 1
November 10, 1982.
Instruction CP-El-4.0-36 identifies the j
daraoing sources as all nco-seismic piping and conduits larger than 2 inches in diameter and non-seismically supported equipment j
ano structures.
The sources are identified by reviewing appli-cable project design documents and/or appropriate markings in the field.
The same criteria epply to identification of targets.
Both sources and targets have been recorded in the Interaction Matrix, which identifies pessible interaction of sources and targets.
Instruction CP-El-4.0-36 also oefines the mode of failure of i
scurces, such as piping, conduits, equipment, and the zone of influence of piping and conduit sources.
It also contains a i
typical Seismic Evaluation Interaction Matrix and a Seismic Inter-acticn Evcluation and Resolution Form, which iaentifies the room where the walk-through has been made and which must be signed by the field walk-through team leader.
The Comanche Peak Seismic Interaction Criteria describe the dynamic impact damage criteria, the failure mode, and the effects at.alysis and source stress analysis for piping, conduits, and hangers.
They also describe target stress analysis for the source and target hanger and the equipment support.
As a result of the TRT finding reported in SSER No. 8, C/S Category 14, TV Electric is conducting a reevaluation of the damage study, which may result in the redefinition of the zore of influence of passible sources.
1 3.
Conclusion and Staff Position:
The issues associated with this allegation are similar to those raised related to C/S Category 14, presented in SSER No. 8.
The actions required by TV Electric associated with C/S Category 14 should encompass this allegation.. -.
4.
Open Issues:
TV Electric shculd identify actions taken or needed to be taken to assure that this concern is beino covered with respect to actions identified in SSER No. 8, C/S Category 14 Reference Documents:
1.
Control of Seismic anc hon-Seismic Components Interaction Evaluations,"
Instructic, CP-El-4.0-36. Texas Utilities Services, Inc., Revision 3, deted December 9, 1983, 2.
Final Safety Analysis P.eport, Comanche Peak Steam Electric Statiun, Section 3.78.3.5.
3.
Standard Review Plan (NUREG-0800), Section 3.7.2, Revision 1, dated July 1981.
4.
Comanche Peak Seismic Interaction Criteria, Revision 1, dated November 10, 1982.
5.
Twenty Damage Study Reports as listea below:
Building (Room / Elevation)
Safeguard Containnent Fuel Handling Auxiliary 58/773'-0" 153/?83'-7" 272/869'-0" 241/852'-6" 61/773'-0" 154/808'-0" 255/810'-6" 219/831'-6" 4
68/790'-6" 155A/881'-6" 253/810'-6" 165/790'-6"79-790'-6" 163/860'-0" 247A/802'-6" 117/792'-0" l
99A/844'-0" 161A/852'-0" 248/810'-6" 115A/778'-0" j
6.
Comanche Feak Steam Electric Station, Seismic Interaction Study Report (November 1982 - January 1983).
1 7.
Safety Evaluation Report, NUREG-0797, Supplerrent No. 8, Civil / Structural Category 14, Control Recm Area.
8.
TU Electric Progrce Plan and Issue-Specific Action Plans, dated October 8, i
1984, Item No. II d.
l C.
Structural Integrity of Cable Tray Supports _(Allegation t' umber:
AC-70, AC-76, AC-77, AC-78, and AC-79) 2.
Characterization:
It was alleged that the following inadequacies occurred in the design of the cable tray supports:
a.
When the equavalent static load method was used in seismic design calculations, a 1.5 factor for seismic acceleration may not have been applied (AC-70).
b.
The damping valves used in the design of cable trays were not consistent with those specified in the Firal Safety Analysis Repcrt
]
(FSAR) (AC-76),
c.
Safety factors for Hilti-bolts were below those reconnended by the manufacturer and by NRC criteria (AC-77).
d.
Holes orilled through the flanges of the channels for the cc tray supports that reduce the section mcdulus were not accounted for in the design calculations (AC-78).
i e.
The allowable stresses for cable tray supports in the Containment Building were permitted to go above the yield stress (AC-79).
l 2.
Assessrrent of Safety Significance:
As a result of the above allega-tions, as well as other cable tray and conduit issues raised by Cygna in its independent Assessment Program (IAP) of the Comanche Peak Project, the applicant has undertaken a requalificaticn program for cable tray and corduit supports.
The deteils of this program are described in DSAP VIII of the Comanche Peak Response Team (CPRT)
Prcgram Plan.
l 3.
Conclusion and Staff Position:
The applicant has developed DSAP VIII, which is inter.ded to be a requalification of cable tray and conduit supports. As part of this program, the applicant is expected to track and resolve all cable tray issues.
4.
Open Issues:
TU Electric should identify actions taken or needed to be taken to assure that the above concerns are addressed as part of CSAP VIII ar.o the Corrective Action Program for Cable Tray Hangers.
Reference Documents:
l
- 1. Transcript of meeting with alleger on llovember 7,1984, pp.115-125,
- 2. Transcript of treeting with alleger on March 23, 1984, pp. 196-213, pp.
216-218.
- 3. Letter, N. H. Williams (Cygna) to J. W. Beck (TUGC0), "Review Issues List Transmittal," dated April 4, 1985.
- 4. t;PC memorandum. R. E. Lipinski (TRT) and J. Devers (TRT) to File, "Site Visit by the TRT Civil /5tructural Grcup " dated March 5,1985.
- 5. Letter, J. B. George (TUGCO) to 8. J. Youngblood (NRC), "Clarification to SGEB Additional Question Responses," dated October 26, 1984. _
- 6. Final Safety Analysis Report (FSAR) for the Comanche Peak Steam Electric Station Units 1 and 2 Section 3.78.
- 7. NRC/NRR Question 130.39,
- 8. Letter, J. B. George (TUGCO) to B. J. Youngbloco (NRC), dated October 26, 1984 (with enclosures)
- 9. Bechtel Power Corp. Design Guide C2.7. "Seismic Category 1 Cable Tray and Conduit Raceway Support System," dated June 1979.
- 10. URS/ John A. Blume & Associates, Engineers, "Analytical Techniques, Models, and Seismic Evaluation of Electrical Raceway Systems," August 1963.
- 11. URS/ John A. Blume & Associates, Engineers, "Shaking-Table Testing for Seismic Evaluation of Electrical Raceway Systems," April 1983.
- 12. Draft Evaluation of the SEP Owners Group Seismic Evaluation of Cable Trays, dated July 5, 1904.
- 13. Code of Federal Regulations 10, Part 50, September 28, 1984.
- 14. Brown & Root, Inc., Purchase Order No. 35-1195-1S837, dated March 2, 1978.
- 15. Letter Wayne Shahan (hilti Fastening Systems, Inc.) to Bob Banks (Brown &
Root, Inc.), dated February 7, 1978.
- 16. Hilti Fastening Systems, Inc. Catalog fH-3908, dated April 1978.
- 17. TL'GC0 Instruction QI-QP-11.2-1, Rev.17 "Instellation of 'Hilti' Drilled-in Bolts," dated December 5, 1984
- 18. Brown & Root, Inc. CPSES Instruction No. 35-1195 CEl-20 "Installation of
'Hilti' Kwik-Bolt.
- 19. Onsite Test Program, "Hilti Kwik-Bolt Expansion Anchors, Brown & Root, Inc.
Comanche Peak Steam Electric, Glen Rose, Texas," April 27 and 28, 1978.
- 20. Brown & Root, Inc. Quality Assurance Receiving Inspection Report dated tierch 20, 1978.
- 21. Quality Control Recertificatien Records for QC Technician level 1 for Hilti Expansion Anchor Bolt Installation for: Donald G. Bishop (February 18, 1985); Phyllis May (August 6,1984); Ken Gouser / August 6,1964); Douglas McCallum (January 24,1985); and Jon Smeat (Deceder 13,1984).
- 22. Testimony of Nancy Williams in Response to CASE r.uestions of February 22, 1984, to CYGNA Energy Services, dated April 12, 1984.
1
- 23. CASE Exhibit C89, Gibbs & Hill, Inc. Job No. 2323, Calculation Set j
SCS-1010, Set #1, Sh. 128 through 146, dated October 26, 1983.
j
- 24. Gibbs & Hill, Inc., "Basis for Using a Minimum Factor Safety Equal to Three for Hilti Expansion Eolts Under SSE Loading for the Electrical Raceway Support System," undated.
- 25. Gibbs & Hill, Inc., Specif.ication No. 2323-SS-30, Appendix 2, "Design Criteria for Hilti Kwik-Bolts " undated.
- 26. Teledyne Engineering Services, "Summary Report Generic Response to USNRC IE Bulletin No. 79-02, Base Plate /Ccrcrete Expansion Anchor Bolts, Pev.
1,"
l August 30, 1979,
- 27. huREG/CR-2137, "Realistic Sesimic Design Margins of Purrps, Valves, and Piping, Appendix B," May 1981.
2P Letter, N. H. Williams (Cygna) to J. W. Beck (TUGCO), "Review Issues list Transmittal," dated April 4, 1985. l
=
D.
Steam Generator Upper Lateral Supports (Allegation humber: AC-81) 1.
Characterization:
CASE has alleged that 1U Electric's original design and analysis, as well as subsequent analyses undertaken in response to an Atomic Safety and Licensing Board's (ASLB) order, failed to demonstrate the adequacy of the lateral restraint beams and the associ6ted reinforced concrete walls in the steam generator compa rttrents.
2.
Assessment of Safety Sicnificance:
In its orignial allegation, CASE challenged the adequacy of the upper lateral restraint bearrs (ULR6) and the associated reinforced concrete walls. Specifically, CASE raised two concerns:
(1) whether the ULRBs were adequately designed and (?) whether the stresses in the steam generctor ccm-partment walls caused by the thermal expansion of the lateral restraint beams uncer design basis accident conditions were within allowable stress limits.
It should be noted that TU Electric had originally neglected the thermal loads in the load combinations used for the analysis. According to TV Electric (see Affidavit of Dr. Robert C. Iotti, May 20,1984), this exclusion of the thermal loads was justified in view of ASME Boiler and Pressure Vessel Code Criteria NF-3231.1(a),(b), and (c), and in the Final Safety Analysis Peport (FSAR) requirement in Section 3.8.3.3.3, ?(b) which states: "Thermdl loads are neglected when they are secondary uno self-limiting in nature and when the material is ductile."
It was, however, not shewn in the affidavit if the conditions for neglecting the thermal loads were, indeed, satisfied.
After reviewing the original analysis performed by TV Electric the allegations made by CASE, and the analysis undertaken by the staff at the ASLB's request, the ASLB concluded in its Memorandum and Orcer (Guality Assurance anc Lesign), December 28, 1983, that "In the face of the possibly conflicting engineering viewpoints of three different parties, we cenclude that applicant has not demonstrated the adequacy of its analysis of the ULRB."
In response to the ASLB's order. TU Electric submitted a plan on February 3, 1984, to resolve this issue by performing further analyses.
The analyses were carried out to:
"Provide evidence that the design for the ULRB and adjacent walls is adequate to withstand mechanical and thermal loacs in a loss-of-ccolant-accident (LOCA) environment.
This will include performance of an analysis to confirm that the forces transmitted to the concrete by the expanding restraint are well within the capacity of the concrete to permit the con-tinued performance by the ccncrete and the support of their intended functions.
This also will include the performance of analyses of the time differential between the peak mech-anical and thermal loads and the realistic stiffness values for the walls."
The analysis procedures and the results thereof, were presented in an afficavit by Dr. Robert C. Iotti on May 20, 1984. _ _ - ___ _ ___- _-___ _ ___ - __-_
0 e
In its affidavit of August 26, 1984, "Partial Answer to Applicant's Stateaent of Material Facts Regarding the Upper Lateral Restraint Beam," CASE raiseo several questions regarding the validity of TV Electric's analyses.
Specifically, CASE was concerned abuut the inclusion of the lower restraint beams in the analyses; the effects of LOCA loads und of shear stresses in the concrete, bolts anc bcams; and the conservatism of the various assumptions used in the unalyses.
On August 14 and 15, 1984, the NRC and Brookhaven National Laboratory (BNL) conducted an audit at the offices of Gibbs and Hill, Inc.
(G&H) in New York, New York on the subject of the ULRB for the steam generator compartment. At this audit, various aspects of the structural analysis and design of the steel beams and concrete walls were reviewed.
Based on this review, several items were identified for further investigation and verification.
These involved two major issues, namely, treatment of seismic loads ano verification of modified NASTRAN cencrete analyses, and a number of secondary issues dealitig with the assumptions that were used in the structural analysis.
Following the investigation of these issues, BNL also performed an independent engineering evaluation of the ULRB and the associated reinforced ccncrete walls. A report entitled, "BNL Review of Texas Utilities Generating Company Comanche Peak Steam Electric Station Upper Lateral Restraint Beam - Steam Generator," was submitted to hRC on January 10, 1985.
After the audit of August 14 and 15, 1984, Ebasco Services. Inc. per-formed the evaluation of several check problems proposed by BNL in order to verify Ebasco's modifications to the NASTPAN Code.
The medi-fications were inserted into the code to perform concrete structural analysis with cracking.
The results wer( sent to Bhl early in September 1984.
Af ter studying the output ano af ter ccnsultation with hRC staff, it was decideo to further investigate one of these problems in greater detail. The specific problem (identified as CA29 in the Ebasco submittal) involves the prediction of load / deflection / cracking behavior and failure conditions for a reinforced concrete beam rigidly held at both ends and centrally loaded, i
A deteiled finite element grid consisting of 200 elements,10 layers through the thickness and 20 civisions along the half length was developed by Ebasco.
BNL used the same mesh and boundary conditions and ran a comparable analysis using the NFAP program (developed at BNL).
The results of these two runs can be described as follows:
(1) At low loads (about 15 kips in the linear region, before the development of substantial concrete elentent crackine), the two runs yield comparable load-deflection output.
(2) At higher loads (about 22 to 24 kips), the BNL/NFAP results predict diagonal tcnsion cracking failure occurring near the supports for the beam.
This result is very close to the value of ultimate capacity predicted by Zsutty's ecuation, which is based on statistical evaluation of extensive experimentally determined data points (Bazant & Kim, paper 81-38, American Concrete Institute (ACI) Journal September - October 1984).
It should be noted that this failure load could also be deduced from the ACI Code.
(3) The failure pattern predicted by NFAP indicates that multiple cracking will develop in some of the elerrents of the mesh, prior to failure of the beam.
(4) The results from the modified NASTRAN indicate no failure, even at a load of 32 kips where the run was terminated.
(5) The predicted crack pattern for this NASTRAN run (Septerrber 1984) did not indicate any trultiple cracking.
In fact, the brief review of the output from all four sample problems presented by Ebasco did not show any multiple crack patterns.
At this stage, a second meeting was held between BNL and Ebasco to discuss these ccmparisons.
This meeting was attended by R. Iotti and H. Chang who represented Ebasco/G&H and S. Sharma and M. Reich from BNL, The outcome of this meeting was that Ebasco would check the NASTRAN results ano screen the problem to vali'date the Ebasco predicted results. At a subsequent meeting (November 5,1984) at BNL, the results of the new computer runs for Ebasco CA29 were presented by H. Chang.
These results can be sunearized as follows:
(a) Ebasco was able to show a failure load of 24 kips (compar-able to test data) cnly by assuming a tensile strength of 120 psi, which is four times lower than the strength that would normally be assurred for the concrete.
(It should be noted that in the September run, the concrete tensile strength used in NASTRAN was 546 psi.)
(b) The Ebasco results showed that fcilure is associated with shear failure accompanied by very large displacements. This is contrary to the results of other studies, including the ENL/NFAP results, as well as experin.ents.
Diagonal tenstori failure in concrete is typically a brittle failure at normal displacements.
(c) The results of the Ebasco runs indicated that the assumed value of the shear retentien factor has no impact on the computad failure loads (this factor was varied from 0.2 to 0.4).
This is not supported from studies reported in the literature, nor from the BNL/NFAP results. An increase in the shear retentien factor from 0.2 to 0.4 should lead to an increase in ultimate capacity of the beam of about 20%.
(d) The trultiple crack pattern shown by Ebasco for this new run indicated cracks orthogonal to each other.
This result would not normally be expected and needs further explanation.
On the basis of these results, PNL concluded that the Ebasco formulation of the ccncrete cracking tredel as implemented in their NASTRAN version led to incorrect results for this veri-fication problem.
Therefore, the c.0%uscy of the results from the code for the steam generator compartment cracking prediction are questionable.
As part of Mechanical and Piping Category 18, "Equipment and Pipe Whip Restraint Bolt and Nut Problems," the TRT investigated allegations related to the improper shortening of anchor bolts in the steam generator upper lateral supports. As a result of j
that review, the TRT requested TU Electric to provide evidence, such as ultrasonic measurement results, to verify acceptable bolt length (letter, from D. G. Eisenhut to M. D. Spence, November 29. 1984).
In response to this request, this issue was included as Item V.b of the Comanche Peak Response Team (CPRT) Program Plan.
TU Electric has acknowledged that the results from ultrasonic tests to determine bolt length in-dicateo insufficient thread engaaement in specific locations andhasproposedanactionplan(V.b)toaddressthisissue.
In additicn, TV Electric has included a review of the overall design adequacy of both the upper and lower steam generator lateral restraints for all load combinations as part of the CPRT Civil / Structural Discipline Specific Action Plan (Item VIII).
The components to be reviewed include the teams, the anchorage of the beams to the concrete walls and the walls j
themselves.
l 3.
Conclusion and Staff Position:
The NRC has not concluded its investigation of this allegation. The overall design adequacy of both the upper and lower steam generator lateral restraints for all load ccmbinations remains to be addressed.
As noted in Section C.2, the calculations presented for the non-linear concrete cracking of the Steam Cenerator Compartment are considerea unverified due to questions arisino from the sample problem CA29 output from the tredified Ebasco/NASTRAN Code, j
Therefore, any further use of the results of this code in the evaluatien of the steam generator compartment walls should be carefully assessed.
4.
Open Issues:
TV Electric should identify actions taken or needed to be taken to assure the overell design adequacy of both the upper and lower steam aenerator lateral restraints for all loao i
cctbinatiers Re'Terence Documents:
1.
Affidavit of Robert C. Iotti Regarding Upper Lateral Restraint Beam Before the Atomic Safety and Licensing Board, May 20, 1984 - Comanche Peak Steam Electric Station, Units 1 and 2, Docket Hos. 50-445 and 50-446.
2.
Applicant's Motion for Sunrnary Disposition, May 20, 1984.
3.
Applicant's Reply to CASE's Answer to Applicant's Motion for Summary Disposition Regarding the Upper Lateral Restraint Beam Before the Atomic 1
Safety and Licensing Board, October 26, 1984.
4.
Minutes of Meeting on August 14 and 15, 1984, of the audit conducted by NRC and BNL at the offices of Gibbs and Hill, Inc., New Ycrk, New York.
5.
BNL Review of Texas Utilities Generating Company Comanche Peak Steam Electric Station Upper lateral Restraint Beam-Steam Generator, January 10, 1985...
E.
Voids in Concrete Walls and Slabs (Allegation Number: AC-55) 1.
Characterization:
It was alleged that voids and concrete of poor quality were encountered.
2.
Assessment of Safety Sionificance:
The hRC Technical Review Team (TRT) became aware of the allegatiens during an interview with the alleger in March 1985.
The alleger stated that he was responsible for drilling cores with diameters ranging from 1-1/2 inches to 24 inches to provide wall and floor penetrations for items such as pipes and electrical ccnduits.
He further stated that at tiros core j
recovery was very poor.
During various parts of the conversation, j
he estimated the incidence of poor recovery at values ranging from 12 to 18 percent.
In response to a question, he stated that poor j
recovery was experienced for all ccre diameters.
He was vre.bie to ioentify any specific locations where recovery was poor.
The 1RT noted that core barrels with the primary purpose of drilling holes differ from those intended to salvage cores as test specimens; therefore, relatively poor core recovery may be anticipated in hole drilling cperations, particularly for the smaller diameters.
How-ever, the fact that poor recovery was alleged to have cccurreo for both large diameters and small diameters and that goed core recovery was experienced over 80 percent of the tirre convinced the TRT that the allegation of poor recovery could not be dismissed as an arti-fact of core drilling without further investigation, i
In its investigation of the records, the TRT examined over 200 Component Modification Cards (CMCs) releting to the cutting of reinforcing steel in the Fuel Handling Stilding, Auxiliary Building i
and Containment Buildings.
For each requested hole, meticulous attention was given to location of reinforcing bars.
It was also noted that drilling could proceed only af ter permission from the designers, which included a stipulation as to which bars, if any, could be cut.
However, there was no systematic recoroing of the condition of concrete encountered during the drilling operation.
There were 20 nonconformance reports (NCRs) cealing with concrete damage associated with or discovered by core drilling; 15 were concerned with drilling for Hilti bolt installations, and of these, 10 were for problems of spalled concrete caused by the drilling operation itself.
Of the five NCRs associated with core drilling operations, three uncovered pre-existing voids. While this evidence illustrates that job persorr.el discovered and corrected some de-ficiencies indicated by poor core recovery, the number of such events is not comparable to the number suggested by the alleger.
The TRT examined available core barrels at the project and discovered that there never was a core barrel with a diameter exceedino 17 inches.
Thus, the allegation is partly in error in stating that holes up to 24 inches were orilled.
3.
Conclusion and Staff position:
The TRT could neither substantiate nor refute this allegation.
Since the TRT cannot conclude from available data whether poorly consolidated concrete constitutes a.
~
significe;,t problem, some investigation of the concrete in place may need to be undertaken Iby TV Electric.
4.
Open Issues:
TV Electric should identify actions taken or needed to be taken to address the concern of poorly consolidated concrete.
Reference Documen_ts :
1.
Component Modification Cards for Reinforcing Steel Cutting in Fuel Handling Building, Auxiliary Building, and Containments.
2.
lwenty Nonconformance Reports (NCRs).
1 i
l.
,,c
. - - -., - -.e
F.
Reactor Cavity Liner Plates {Allecation Numbers:
AC-82,AC-84) 1.
Characterization:
a.
Regarding the stainless steel liner for the reactor refuelirg cavities in Units 1 and 2, it was alleged that: (1) the floor and wall liner plates were improperly located; (2) the tolerance on the gap between plates, the elevation between floor plates, and the overlap of the corner angle and the floor plates were not maintained; (3) improper or uncontrolled aaterials were used; (4) the helson anchor studs were improperly located and welded;-(5) a "leak-chase" channel is missing; and (6) voids exist in the concrete behind the liner (AC-82),
b.
Allegation AC-84 repeats that portion of AC-82 dealing with improper installation of floor plates.
2.
Assessment of Safety Sianificance AC-82:
The NRC Technical Review Team (TRT) interviewed the alleger in March 1985.
The alleger also accompanied the TRT on an inspection tour of the Unit 1 ano Unit 2 reactor refueling cavities, lhe alleger stated that he had been involved in the fabrication of all the liners.
In his opinion, the tolerances and practices used in construction did not conform to the design specifications.
He further stated that he could identify tre location where the "leak-chase" channel was omitted and coule locate specific examples of tolerance deviations and poor workmanship in the field.
Members of the TRT and the alleger conducted an inspection of the Unit 1 and Unit 2 reactor refueling cavities in Farch 1985.
As examples of poor workmanship in Unit 1, the alleger pointed out a wall liner weld showing a localized rust discoloration, gouge marks on the walls, dirt accumulations under bolted flanges, and loose and dirt-encrusted shims between flanges.
As an indication of potential veids in the concrete behind the liner, the alleger pointed cut a depression in the wall plate of the deepest Unit 1 pool.
The depression was approximately 1 foot in diameter, deflected inward towards the concrete, cod emitted a hollow sound when tapped.
In other sections of the Unit I refueling cavities, hollow sounds were emitted frcm the wall liner plates when they were tapped at the mid-span between welds.
In the vicinity of the welds, the emitted scunds were solid. The alleger also pointeo out what he considered to be general examples of poor alignment ano fitup of the Unit 1 floor and wall plates.
The location where the "leak-chase" chanrel was alleged to have been omitted was not identified.
At the time of inspection of the Unit 2 reactor cavity, TV Electric was in the process of inspecting suspected voids behind the wall liner plates and filling them as necessary. This was the subject of a previous allegation, and is discussed in SSER ho, 8, Civil /
Structural (C/S) Category 4.
Towards this end, inspection holes were drilled throuah the liner and the void extent determined, lf necessary, grout was injected through the holes to fill the void, and the liner w6s then repcired.
The TRT inspection of the Unit 2 liner centered on these activities.
The inspection holes that remained open did not exhibit any signs of voids; however, it could be observed that there was a gap between the concrete and the lirer.
The general appearance of the Unit 2 liner seco.ed better than that of the Unit I liner.
The allegations may be grouped into three addressable issues, first, were the Unit 1 and Unit 2 cavity liners fabricated in e tashion consistent with TV Electric's connitments and the applicable safety standards? Second, do voids exist behind the liners and, if so, do they have an impact en safety? And, lastly, was a "leak-chase" channel omitted and, if so, whet is the safety significance of the cmission?
Allegations have been made previously concerning the fabrication of j
the reactor cavity liner.
These allegaticns, which are discussed in SSER ho. 10, Mechanical and Piping (M/P) Category 43 (Allegations AM-11, AW-40, AW-42, AQW-80 and AQW-81) are similar and, in some cases, identical to the fabrication allegations considered here.
In SSER No.10, the applicable functional requirements, fabrica-tion techniques, inspection procedures and docunentation for the liner were reviewed and assessed.
(Also, see SSER No. 11, QA/QC Category 6.)
In the case of the functional requirerrents, the previous review concluded that the primary purpose of the plate is to provide a smoother, less perrr.eable surf ace, which is easy to decontaminate j
and which provices a construction form for the cavity, but does not provide any structural integrity to the con &ete structure.
Legarding the fabricaticn and instellation procedures, the previous review deterrnined that the Brown & Root procedures for the fabrication and installation of the lirer, including the welding, are consistent with Gibbs & Hill specifications.
The TRT review and inspection fcund no safety significance for the specific allegations concerning incorrect welding, poor fitup, rust stained welds, plate mislocations, and floor plate overlap tolerances.
A nurrber of other allegations citing voids in concrete have been niade and are under review. One of these allegations, SSER No. 8, C/S Category 4, AC-25, involves voids in the concrete behind the stainless steel liner of the Unit 2 reactor cavity; a situation similar to this case.
This allegation was substantiated by TUEC actions to investigate and repair suspected voids behind the Unit 2 cavity liners.
Investigation and repair were underway at the tirre of the TRT review.
The depressea plate and hollow
. a sounds on tapping indicate that voids may exist behind the Unit I cavity liner. Given these indications and the substantiated findings for Unit 2, an inspection for voids behind the Unit i reactor cavity liner plate is warranted.
The last issue concerns the "leak-chase" channel, which allegedly was omitted from a weld in the reactor cavity liners. The i
location of this omitted component was not identified.
Since j
these components are behind the liner plates and under concrete, the allegation cannot be verified visually.
The "leak-chase" channels form a leak detection and leakage water recycling system for the liner. However, the NRC staff position is that the liner plate itself is not safety-related; therefore, systems desianed to monitor their integrity are not safety-related.
If a single "leak-chase" channel is missing, it will compromise, to some extent, the capacity of the leak detection system to perform its function.
1 If a leak develops in the affected weld, the system operators will have no indication of the source of the leak.
If leakage is sufficient to require repair, potentially all of the welds below the water line may have to be inspected.
In effect, the "leak-chase" system is installed fnr the convenience of the applicant, and any omissions in that system will pose operational, but not safety-related, problems.
The "leak-chase" chanrel issue is similar to that raised in allega-tion ACW-82, SSER No.10, M/P Category 43, concerning an allegedly defective block in the floor liner "leak-chase" system.
In that instance the allegation could not be sut,stantiated and was deter-mined to have no safety significance.
AC-84 Allegation AC-84 is dealt with in the above discussion.
3.
Conclusions and Staff Positions:
AC-8?
The issues related to the fabrication practices and quality of workmanship used in the construction of the reactor cavity liners were addressed by the TRT in SSER No. 10, M/P Category 43.
Based on that effort, the TRT determined that the above allega-1 tions cencerning fabrication and worknanship issues are not of safety concern.
In the inspection of the Unit I reactor cavities on March 7, 1985, the TRT noted some indications of potential voids in the concrete behind the cavity liner plates.
Further, TV Electric has uncovered veids in the concrete behind the Unit 2 cavity liner plate. Given these observations, there remains some question regarding the itegrity of the concrete behind the cavity liner in Unit 1.
s The TRT could neither substantitato ror refute the allegation con-
~
cerning the omission of a "leak-chase" channel from a weld in the reactor cavity liner. However, since the liner plate itself is not safety related and the leak detection system is installed for the applicant's convenience, the TRT has determined that the emission, if real, is not a safety concern.
AC-84 Allegation AC-84 is dealt with in the above discussion.
4.
Open Issues: TU Electric should identify actions taken or needed to be taken t'o verify the integrity of the concrete behind the cavity liner plates in Unit 1.
Reference Documents:
1.
Safety Evaluation Report, NUREG-0797, Supplement No. 10, April 1985.
2.
Safety Evaluation Report, NUREG-0797, Supplement No. 8.
3.
Memorandum for V. S. Necnan from 0. D. Parr, "Staff Position en Liner Plate for the Refueling Cavity, Fuel Transfer Canal ana Fuel Transfer Tube,"
dated February 26, 1965.
1 1
l
. i
G.
Containment Liner Plate (Allecation Numbers: AC-80,AC-85) 1.
Characterization:
During a meeting between the NRC Technical Peview Tecm (TRT) and the Citizens Association for Sound Energy (CASE) on November 7,1984, it was alleged that a plate attached to the containment liner, which serves as the support point of a pipe support, was stressed to 100 ksi (AC-80).
During an Atomic Safety Licensing Board (ASLB) hearing (transcript dated July 26,1982, page 2847, line 17 to pace 2851, line 5), it was alleged that the containment liner plate is out of plumb by 4 inches (AC-85).
2.
Assessment of Safety Sionificance:
AC-80 The alleger stated that while he was employeo at Comanche Peak he was a member of the STRUDL Group analyzing pipe supports.
He i
recounted an instance where he includeo into his pipe support model a cantilevered plate which was attached to the containtrent liner plate and which served as the support point for a pipe i
support.
The results of his analysis showed that the cantilevered plate was stressed to 100 ksi (AC-80), well beyeno the yield strength of the plate.
The alleger inforrred Nuclear Power Services international (NPSI), the pipe support designer, of the overstressed conoition.
NPSI, in turn, informed Gibh & Hill (G&H), the desigr.ers i
of the cantilevered plate.
Acenrding to the alleger, G5H replied h t the cantilevered plate was part of the design of the liner plate and not to model it any more.
l The TRT cuestioneo the alleger in an attempt to obtain trore specific information as to support location, elevation, and line number.
The allager could not be specific, but thought the support was on the centainment spray system and that the support could be observed from the uppermost operatint] floor of the Peactor Building.
He further c
stated that the cantilevered _ plate had a piece of tube steel attached to its top and bottom. These pieces of tube steel were part of the pipe support desigr.ed by NPSI.
The TRT inspected the piping observable from the uppermost crerating floor in the Unit 1 Containment Buildirg, with emphasis on the containment spray system.
The TRT found that the pipe supports for the contairr.ent spray system were attached to the liner plate by built-up beam sections.
The TRT obtained drawings which shewed the details of the built-up sections that had been observed by the TRT in the field.
On March 6, 1985, the TRT conductad another field inspection of the pipino system observable from the uppermost operating level of the Unit 1 Contaiscent Building, and again did not observe any attach-.
ments of the type described by the alleger. The TRT also discussed thi:; issue with a TV Electric representative, who recalled an instance where an individual analyzing pipe supports incorrectly modeled a cantilevered plate which showed the plate to be over-stressed.
However, he could not recall the specific support in which this occurred.
The TRT met again with the alleger on March 23, 1985, in an attempt to obtain n. ore specific information.
The TRT explained that cbservaticn showed the containment spray system to be supported by built-up sections off the lirer plate rather than a single cantilevered plate.
The alleger indicated that this was not the type of support he was analyzing and offered to draw a sketch of the configuration.
The sketch essentially consisteo of a horizontal plate cantilevered off the liner plate with a piece of tube steel attached to the top and bottom. Upon further cuestioning, the alleger stated that this support might have been associated with the heating and ventilating system rather than the containment spray i
system, but recalled the line number becan with a "V."
As a followup to the March 23, 1985, meeting, CASE supplied addi-tioral information regarding the subject plate in a letter dated April 12, 1985. Attached to the letter was a photograph depicting the containment spray system (CT) and the hydrogen purge exhaust system (VA) on which the allecer had indicated the pipe supports associoted with the CT and VA systems by arrows. On July 25, 1955, the TRT held a telephere conference with the alleger to clarify this apparent aisunderstanding.
It was agreed that the TRT would furnish the drawings associated with the CT and VA systems to the alleger to enable him to pinpoint the subject plate.
On January 23, 1986, the TRT met with tre alleger to discuss this matter further.
The TRT ano the alleger revieweo pipe support drawing VA-1-005-021-C82R for the hydrogen purge exhaust system.
The alleger could not identify this support as the particular one he analyzed and found to have high stresses, but indicated that the connection detail to the liner plate was similar.
This connection consists of a 1/4-inch plate welded to ano cantilevered off of the liner.
Two tube steel members are positicned above and below the plate, sandwiching it between them, and are attached to it with three bolt connections. The two tube steel members form the basc for the pipe support, which is raised from them and attached to the pipe.
The drawing indicated that the bolt holes are oversizeo and clearances are provided between the tube steel members and the plate so that hard contact is not made between these surfaces.
Owing to these clearances, this pipe support nenber is unusual in that it is not rigidly fixed to the liner plate, its foundation, or in any direction and may be expected to exhibit displacements relative to the liner.
The alleger remembered that the support he analyzed had a gap between the tube steel members and the liner plate.
The alleger expressed concerns with the connection detail including possible slippage of the tube steel members away from the liner plate due to the oversized hcle in the cantilevered plate, the grade of materiel - - _ _
used for the bolts tying the tube steel n:tmbers to the 1/4-inch plete, the analysis methods used to cutlify the connection, and whether the vendor-certified, as-built drawing accurately reflects the as-built condition.
AC-85 In assessing the allegation that the containment liner plate is out of plum by 4 inches, the TR1 reviewed the dimension 61 tolerance requirements for the liner and also the fielo measurerents taken to assure that these tolerances were met. The centainmert wall liner consists of 19 circumferential rings, each approximately 10 feet in depth, welded together.
The liner was used as the interior formwork during the placement of the concrete for the centainment walls.
Brcwn & Root (B&R) construction procedure CCP-34 was used to per-form the dimensional inspection of the liner plate.
The procedure required'thet the plumbness of the liner be determined at 30 deuree intervals for each circunferential segment.
Vertical plumb was to be determinea both before and after concrete placement.
Given these two rcquirements and a total of 19 circumferential rings per i
wall, there would be a total of 456 plunbness measurements required per containment structure.
Section 4.2.3 of the procedure outlined the tolerances for vertical plumbness before concrete placement.
There were three dir.ensional requirements that had to be met prior i
to corerete placement:
(1) the liner had to be plumb within 24-11/16 inchen at any height of the liner, measured from a theoretical line extending up from the bMe of the liner; j
(2) cut-of-plumb could not exceed 2 inches horizontal in any 15 i
feet vertical; and (3) the rate of change of plumb within a plate section coulo not exceed 1 inch horizontcl in 8 feet vertical.
The measurement for the rate of change of plun,b within a plate section was not to be made closer than 12 inches from a welded seam.
Section 4.6 of procedure CCP-34 outlined the requireo tolerances after the concrete was placed against the liner. The above i
tolerances were to be met except that the liner in the as-built condition had to be plumb to within t5 inches at any height of the liner when measured from the base.
Therefore, if the allegation that the liner was out of plumb by 4 inches was correct, it would not constitute a violation of the dimensional tolerance requirements.
In order to determine if, indeed, the liner was out of plumb by 4 inches, the TRT reviewed the plumbness ceasurements recorded on form BRDR-2, as required by procedure CCP-34 The TRT found that for Unit 1 the largest amount the liner was out of plumb after concrete placen:ent was 2-5/8 inches.
The Unit 2 containment liner was found to be a maximum of 2-1/16 inches out of plumb after ccncrete placement.
Therefore, neither the Unit 1 nor the Unit 2 containment liner is out of plumb by 4 inches, l
J 3.
Cericlusion and Staff Positions:
AC-80 The TRT reviewed the details shown on pipe support drawing VA-1-005-0?l-C82R and agreed that this is an unusual pipe support design that merits further investigation.
AC-85 The TRT determined that there is no structural safety significance associated with this issue since (1) the allegation that the Unit 1 and 2 liners are out of plumb by 4 inches could not be substantiated and (2) even if the allegation were substantiated, the 4-inch value would not violate procedural tolerance requirements.
4.
Open Issues: TU Electric shculd identify actions taken or neeoed to be taken to investigate the adequacy of the unusual desian shown of the pipe support shown on pipe support design drawing VA-1-005-021-C82R.
The investigation should include consideration of other pipe supports that may be attached to the containment liner wall that are similar in design.
Reference Documents:
1.
Dimensional Record Procedure CCP-34.
2.
Dimensional Record Forms ERDR-2, BROR-1, 3.
Drawing CT-1-033-010-C82R.
4.
Drawing CT-1-033-403-C82A.
5.
Drawing 2323-S1-0514, 6.
Drawing VA-1-005-04-C82R.
7.
Crawing VA-1-005-021-C82R.
8.
Letter from CASE to V. Noonan (hRC), dated April 12, 1985.
9.
Transcript of interview with CASE, dated November 7,1984.
10.
Transcript of interview with CASE, dated March 23, 1985 11.
Drawing BRHL-VA-1-RB-003.
12.
Drawing BRHL-CT-1-RB-024, 13.
Transcript of interview with CASE, dated January 23, 1986.
j i
1
.l-H.
In1 proper Construction Practices (Allegation Numbers: AC-83,AC-86) 1.
Characterization:
It was alleged that the following improper construction practices occurred at various times:
a.
Large holes (diameter and depth) were filled with grout when they should have been filled with concrete (AC-83),
b.
Comanche Peak is unsafe and improperly constructed (carpentry, form-work, and steel work) (AC-86).
Allegation AC-63 was taken from a transcript of a closecut inter-view between the NRC Technical Review Team (TRT) and an alleger on Mdrch 6, 1985.
AC-86 was taken from an ASLB hearing transcript oated July 28, 1982 (page 2759, line 20 to page 2761, line 6; page 2788, line 14 to page 2808, line 24; page 2831, line 9 to page 2833, line 14).
2.
Assessment of Safety Significance:
AC-83, The TRT examined Brcwn & Root (BLR) Frocedure CCP-12. "Concrete Patching Finishing and Preparation of Construction Joints," which provides for repair by dry pack, mortar, or concrete. The alleger's mention of "grout" apparently refers to dry pack. Dry pack is a form of grout, which consists of a mixture of cement, seno, and water.
The two terms may be useo interchangeably.
The rerair precedure states in Section 4.1.2.3:
"In general, ory pack patching shall be used to fill cleaned out voids not exceeding approxinately 6 inches in depth or approximately 1 cubic foot in volume.
Larger voids may be dry packed if more practical." Section 4.1.2.5 of the procedure states: "Defective voids having a r.ominal depth exceeding approximately 6 inches shall be repaired by use of j
concrete or dry pack." lhus, while concrete is permitted as a patching material in large voios, it is never required.
Dry pack is en acceptable repair raaterial.
AC-86 l
The allegation of unrafe and improper construction is nonspecific; therefore, there were no specific portions of the project which the TRT could examine to investigate this allegation. However, in reviewing over 80 allegations, which had various degrees of specificity, the TRT examined the records for 600 concrete placements and examined at least half of those placernents in the field.
In all but a very few, the construction was either adequate or was trade adequate by execution and implementation of a nonconformance report.
The few remaining unresolved matters were designated open issues and were turned over to TV Electric for resolution.
In resolving one open issue VU Electric tested the in-place strength of 200 randomly selected placerrents.
All were found to be adequate.
.,o The adequacy of form-work carpentry is important to construction sa fe ty.
It is a matter of inspection, which is occumented on the pour card. However, eventually all form-work is removed and has no influence on the safety of the permanent structure.
The alleg6 tion also mentioned steel placement.
The TRT has investicated several allegations of omitted or misplaced reinforcing steel. Most o' these allegations were accommodated by physically correcting the situation or by calculations demonstrating that the structural component is safe in its as-tuilt condition.
3.
Conclusions and Staff Positions:
AC-84 The TRT considered'that the allegation that grout was used in repairing voids that should have been filled with concrete is without foundation because the procedure governing repair does not require concrete in any void.
AC-86 TRT's investigation did not substantiate the allegation that there is widespread unsafe and frproper construction.
Two types of construction were specifically referred to in the allegation, i.e., form-work carpentry and steel placement.
The former, while generally adequate, has no affect on the structural adequacy of the finished product; however, the TRT believes there is some merit to the general allegation of misplaced steel.
4.
Cren issues:
TV Electric should identify actions taken or planned to be taken to address the general concern of misplaced steel.
This issue has similarity to that preser.ted in Section II. A of this enclosure.
Reference Documents:
1.
Crown & Roct Procedure CCP-12..
- _ _ _ - - _ _ _ _ _ _ _. _ _ - _ _ - _ _ - -