ML20237L493

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Voided Rev 1 to QI-QP-16.0-4, Identification of Class IE Equipment W/Deficient IEEE-323/344 Tedfr Repts
ML20237L493
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 07/22/1981
From: Scott B, Jacqwan Walker
TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20237K807 List: ... further results
References
QI-QP-16.0-4, NUDOCS 8708200237
Download: ML20237L493 (3)
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TEXAS UTILITIES GENERATING CO. NU 8 REVISION PAGE CPSES QI-QP-16.0-4 1 JUL 2 2 1981 1 of 2 I

PREPARED BY: J7 m _ C. _7-22-Of IDENTIFICATION OF CLASS () ~ OATE 1E EQUIPMENT WITH DEFICIENT U 4!EEE*323/ M & 6 ,

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1.0 REFERENCES

1-A CP-QP-16.0, "Nonconfonnances and Deficiencies" ,

I 2.0 GENERAL ,

f0R lif0Et0H OHiJ 2.1 PURPOSE AND, SCOPE

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The purpose 'of this ins t ructi on is to detail the use of the Nonconfonnance Report to identify, trace and close out deficient envi rcrvnental tes t doctnentati on for Class 1E equipment and c onponent s. This instnJction is only applicable to this equipment and purpose.

3.0 INSTRUCTION 3.1 IDENT A 0 F All Cl 1 ipment shall be identified by equipment speci fica n r on the attachment to the NCR.

3.2 TRACKING All equipment received wi th insufficient envi ronmental tes t documentation shall be identified and entered on the attachment t o NCR E-80-0088. This NCR shall be used to track and status the deficient qualifications of Class 1E equipment.

3.3 STATUS INDICATORS Hold tags shal.1 not be applied to equipment wi th deficie'nt environmental qualification.

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The NCR will renain open until Qualification Reports are received I and ap' proved on all Class 1E equipment. ,

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All deficiencies shall be doctsnented in the Nonconformance Report (NCR) per reference 1-A employing the exceptions as detailed in this instruction. _ _ . .

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i IE Bulletin No. 79-14 Revision 1 Date: July 18, 1979 Page 2 of 3 Action to be taken by Licensees and Permit Holders:

All power reactor facility licensees and construction permit holders are requested to verify, unless verified to an equivalent degree within the last 12 months, that the seismic analysis applies to the pctusi configura-tion of safety-related piping systems. The safety"related piping includes

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Seismic Category I systems as defined by Regulatory Guide 1.29, " Seismic Design Classificatic,n," Revision 1, dated August 1,1973 or as defined in the applicable FSAR. The action items that follow apply to all safety-related piping 2-1/2-inches _in diameter and greater and to seismic Category I piping, regardless of size which was dynamically analyzed by computer. For older plants, where Seismic Category I requirements did not exist at the time of licensing, it must be shown that the actual configuration of (H M safety-related systems, utilizing pipion 2-1/2 inches in di.ameter and greater, meets design require- i ments.

Specifically, each licensee is requested to:

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1. Identify inspection elements to be used in verifying that the seismic i analysis input information confctus to the actual configuration of safety- l related systems. For each safety-related system, submit a list of design documents, including title, identification number, revision, and date, which were sources of input information for the seismic analyses. Also submit a description of the seismic analysis input information which is )

contained in each document. Identify systems or portions of systems which are planned to be inspected during each sequential inspection identified in Items 2 and 3. Submit all of this information within 30 days of the date of this belletin.

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2. For portions of systems which are normally accessible *, inspect one system h each set of redundant systems and all nonredundant systens for con-formance to the seismic analysis input information set forth in design l documents. Include in the inspection: pipe run geometry; support and restraint design, locations, function and clearance (including floor and wall penetration); estbedments (excluding those covered in IE Bulletin 79-02); pipe attachments and valve and valve operator locations and weights (excluding those covered in IE Bulletin 79-04).

Within 60 days of the date of this letin, submit a description of the results of this inspection.

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  • Normally accessible refers to those areas of the plant which can be entered during reactor operatien.

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i UNITED STATES NUCLEAR REGULATORY C01911SSION WASHINGTON, D.C. 20555 IE Bulletin No. 79-14 Date: July 2, 1979 Page 1 of 3

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SE1SMICANALYSESFORAS-BUILTSAFETY-RELATEDPIFpGSYSTEMS Description of Circumstances:

Recently two issues were identified which can cause seismic analysis of safety-related piping systercs to yield nonconservative results. One issue involved algebraic summation of loads-in some seismic analyses. This was addressed in show cause orders for Beaver Valley, Fitzpatrick, Maine Yankee and Surry. It was also addressed in IE Bulletin 79-07 which was sent to all power reactor licensees.

The other issue involves the accuracy of the information input for seismic analyses. In this regard, several potentially unconservative factors were discovered and subsequently addressed in IE Bulletin 79-02 (pipe supports) and'79-04 (valve weights). During resolution of these concerns, inspection by IE and by licensees of the as-built configuration of several piping systems revealed a number of nonconformances to design documents which could potentially affect the validity of seismic analyses. Nonconformances are identified in Appendix A to this bulletin. Because apparently significant non-conformances to design documents have occurred in a number of plants, this '

issue is generic.

The staff has determined, where design specifications and drawings' are used to obtain input.information for seismic analysis of safety-related piping systems, that it is essential for these documents to reflect as-built con-figurations. Where subsequent use, damage or modifications affect the con-dition or configuration of safety-related piping systems as described in documents from which seismic analysis input information was obtained, the licensee must consider the need to re-evaluate the seismic analyses to con-sider the as-built configuration.

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IE Bulletin No. 79-14 Date: July 2,1979 Page 2 of 3 Action to be taken by Licensees and Permit Holders:

All power reactor facility licensees and construction permit holders are requested to verify, unless verified to an equivalent degree within the last 12 months, that the seismic analysis applies to the agtual configura-tion of safety-related piping systems. The safgty-related piping includes Seismic Category I systems as defined by Regulatory Guide 1.29, "Seissie Design Classification," Revision 1, dated August 1,1973 or as defined in the applicable FSAR. For older plants, where Seismic Category I require-ments did not exist at the time of licensing, it must be shown that the actual configuration of these safety-related systems meets design require-ments.

Specifically, each licensee is requested to:

1. Identify inspection elements to be used in verifying that the seismic analysis input information conforms to the actual configuration of safety-related systems. For each safety-related system, submit a list of design documents, including title, identification number, revision, and date, which were sources of input information for the seismic analyses. Also submit a description of the seismic analysis input information which is contained in each docament. Identify systems or portions of systems which are planned to be inspected during each sequential inspection identified in Items 2 and 3. Submit all of this information within 30 days of the date of this bulletin.
2. For portions of systems which are normally accessible *, inspect one system in each set of redundant systems and all nonredundant systems for con-formance to the seismic analysis input information set forth in design documents. Include in the inspection: pipe run geometry; support and restraint design, locations, function and clearance (including floor and wall penetration); embedments (excluding those covered in IE Bulletin 79-02); pipe attachments and valve and valve operator locations and weights (excluding those covered in IE Bulletin 79-04).

Within 60 days of the date of this bulletin, submit a description of the results of this inspection. Where nonconformances are found which affect operability of any system, the licensee will expedite completion

. of the inspection described in Item 3

  • Normally accessible refers to those areas of ihe plant which can be entere; during reactor operation.

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IE Bulletin No. 79-14 Date: July 2, 1979 Page 3 of 3

3. In accordance with Item 2, inspect all other normally accessible safety-related systems and all normally inaccessible safety-related systems.

Within 120 days of the date of this bulletin, submit a description of the results of this inspection.

4. If nonconformances are identified: ,.

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b. Submit an evaluation of identified nonconformances on the validity of piping and support analyses as described in the Final Safety Analysis Report (FSAR) or other NRC approved documents. Where you determine that reanalysis is necessary, submit your schedule for:

(I) completing the reanalysis, (II) comparisons of the results to FSAR or other NRC approved acceptance criteria, and (III) submitting descriptions of the results of reanalysis.

c. In lieu of o, submit a schedule for correcting nonconforming systems so that they conform to the design documents. Also submit a descrip-tion of the work required to establish conformance.
d. Revise documents to reflect the as-built conditions in plant, and describe rnessures which are in effect which provide assurance that future modifications of piping systems, including their supports, will be reflected in a timely manner in design documents and the seismic analysis.

Facilities holding a construction permit shall inspect safety-related systems in accordance with Items 2 and 3 and report the results within 120 days.

Reports shall be submitted to the Regional Director with copies to the Director of the Office of Inspection and Enforcement and the Director of the Division of Operating Reactors, Office of Nuclear Reactor Regulation, Washington, D.C. 20555.

. Approved by GAO (R0072); clearance expires 7/31/80. Approval was given under a blanket clearance specifically for generic problems.

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i Addenda 2 to July 23. 1986 Testimony (pagig03_), _

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Thedraftreport(documentla)of 50-445/85-16 and 50-4 /85-13 states the following on page 4-2, "The NRC review of correspondence in. file determined j that TUGCo never adequately responded to all aspects of the IE Bulletin; i.e., ~ t 1 the TUGCo responses did not indicate that all seismic tategory I piping (regardless-of size) would be evaluated to the actiom itemrsin the IE Bulletin: if it was dynamically enalyzed by computer , . . ." ,

l The TUGCo letter TXX-3597, dated December 3, 1982y which this statement refers to states the following: .

5_ g "The scope of the program has been established based upon a detailed review of "

Bulletin requirements and studies initiated to' identify applicable piping  % 1 systems and related components along with the stress analysis problems associated with the subject piping. All related documents pertaining to each stress problem have been identified and the basis for the As-Built Verification Program is firmly established. The following defines the scope of the piping being as-built verified to satisfy requirements:of the Bulletin: ,

3 SAFETY CLASS A \

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1 All sizes 2.3 Largebore(2-1/2"andlarger).

2,3,5 High energy lines over one inch that were * -

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computer analyzed.

2,3,5 'g Designated piping, regardless of size. (up to and including the first anchor or terminal connection)- w that interacts with safety-related large bore pipe. N Note: Class.5 piping has been defined as non-nuclear safety-related lines

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contained in. Seismic Category I structures. 1 Safety (Class 2ofand lines regardless the3analysis small bore (2" used) method and smaller) have been non-high excluded energy ,

from the scope of the formal 79-14. verification program for the following reasons: w 4

1. These lines are analyzed after the piping has been installed and the as-built configuration is known.. -

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2. Support locations used in the analysis are field verified, prior -

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3. Deviations from the 3s-designed support locations are design N reviewed and rer.onciled with the analysis.

'In order to interpret the piping exclusions from IEB 79-14, the following -

clarification from Revision 1 of IEB 79-14 page 2 is necessary:

"The action items that follow apply to all safety-related piping .

2-1/2-inches in diameter and greater and to seismic Category I piping,,

regardless of size which was dynamically analyzed by computer." ~

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,y i ^"TheBulltDnstates"allpcwerreactorfacilitiesandconstructionpermit holders are requested to verify, unless verified to an ecuivalent degree within the last 12 months, that the seismic analysis applies to the actual config0 ration of safety-rtisted piping systems." I believe it was the intent of the 6ulletin that if the utility had an equivalent program in place, that i

the Bulletin didn't require a whole new program for pust efforts. At the time the Bulletin was issued, TUGCo already had a program in plac, that satisfied tht Bulletin for Safety Class 2 and 3 small bore npn 41gh energy lines. These particular lines were field run and support locations were field verified sprior.zto completion of analysis as a part of the already in place (at the 1ime of the Bulletin) construction program. There would appear no need to add i

_, this class of lines to the formal 79-14 verification program.

-%. It;is my understanding that only high energy lines over 1 inch were 1* '

c(2puter analyzed to limit the number of postulated energy breaks, as s discussed in TUGCo internal memo dated August 6, 1982 (attachment 1). Those

< ,, 1 inch and under which were not computer analyzed are excluded by the Bulletin.

I would suggest that for further clarification you talk to John Fair, IE:HQ. There has never been an exception taken to the TUGCo reply by F!rgion IV or the IEB 79-14 lask Force which was compiled of Region IV, s . IE:HQ, and NRR.

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CPP. 4 t.XAS UTILITIES SERVICESM PINC(.'- ? ;

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r IE BULLETIN 79-14' REF: CPP-8001 DATED JUNE 26, 1982 (ATTACHED) ' '

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Technical Services has responded to TUGC0 QA Audit TCP-32 per attached memo CPP 8001 dated June 28, 1982... Deficiency f 3 ef TCP-32 question 6d._

our compliance with 79-14 in the area of small bore piping. As explained in CPP-8031, small bora piping analyzed by the Tomographic method is c extremely conservative, and the as-built piping configuration is known

' prior to perfoming the analysis and subsequent support design. Thist small bore piping is not required per 79-14 and has been excluded from the As-Built Verification Program. However, Class 2 and 3 high energy piping over one inch has been computer analyzed to limit the number of postulated line breaks, and small bore piping falling in this category is included in the scope of As-Built.

Per discussions in Dallas with TUSI Nuclear Licensing personnel David Wade and Bob Dacko, a formal interpretation of 79-14 to be. presented to the NRC is not warranted at this time. Nuclear Licensing reconmends that this response be made upon completion of Program activities.

If any further information is needed, please contect the undersigned. j l

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, As-Built Group Supervisor f

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cc: M. R. McBay J. R. Johnson H. A. Harrison .

HomerSditiidt e David Wade (TUSI - Dallas)

Bob Dacko (TUSI - Dallas)

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Addenda 4 to July 23, 1986 Testimony (page 603)

In the draft report of 50-445/85-16 and 50-446/85-13, pages 4-2 it also states that the NRC review of correspondence detemined that TUGCo never addressed the reporting of nonconfomances.

. . . and nonconformances were not reported to the NRC as required by pages 2 and 3 of the IE Bulletins. IN TUGCo internal. letter CPPAf 24.163 dated October 22, 1982 the Engineering Manager stated that the reporting of nonconformances is an area of the IE Buli ttin which we must take exception and we will not identify nonconforming condition; however this exception was not stated in the final report to the NRC."

The following is the exact quote from the TUGCo internal letter CPPAl24.163 dated October 22, 1982:

" Requirements of the Bulletin which concern reporting of nonconformance items is an area in which we must take exception. We will not identify analyzing the as-built conditions while the plant is undergoing construction. Careful records are maintained on all nonconforming conditions and to date no significant differences between our field design change system and the as-built conditions have been found."

I have the following comments: '

1. The inspector imediately draws the conclusion that TUGCo never reported nonconfomances to the NRC as required by pages 2 and 3 of the IE Bulletin because of the statement in the internal TUGCo letter. As I indicated, this statemuat is in error since it is my understanding that the reporting requirement is met by filing of reports in accordance with 10 CFR 50.55(e) for a construction facility.
2. In the review of IEB 79-14, the reporting requirements for nonconformances appear to have been established for an operating plant where a timely determination of system operability and compliance with Technical Specifications is required. The reporting requirements for a construction facility is not cicar.
3. I have talked to Mr. Homer Schmidt, who was the TUGCo Manager of Licensing at the time of the TUGCo December 3, 1982 reply to IEB 79-14.

The TUGCo internal letter CPPA#24.163 was addressed to Mr. Schmidt. The TUGCo licensing group prepared the IEB 79-14 response. Mr. Schmidt.

indicated that TUGCo licensing had interpreted the 79-14 reporting requirements to be applicable to an operating reactor facility. He stated that careful records of all nonconforming conditions were being maintained and that significant differences were to be reported in accordancewith10CFR50.55(e). Therefore, the paragraph in CPPAf24.163 was not considered appropriate for inclusion in the TUGCo December 3, 1982, reply to IEB 79-14.

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l Addende 1 to July 23, 1906 Testimony (page 583)

Phillips assigned McKluskey to inspect IEB 79-14 without any considerations of past inspections of IEB 79-14. It is my understanding that he assigned McKluskey to inspect this Bulletin because he did not consider that Region IV had looked at hardware associated with IE Bulletis c6rrective action. .The-hardware associated with this Bulletin has, since the beginning, been inspected on numerous occasions, including NRC Inspection Report 50-445/82-05 which looked at the initial program.

In this first inspection of the TUGCo 79-14 verification program, R. Brickley (who was assigned out of the Vendor Branch to the IEB 79-14 Task Force) was on site and looked at pipe support hardware. C. E. Johnson inspected the TUGCo IEB 79-14 verification program.

The Special Inspection Team (SIT) was a full team inspection devoted to pipe and pipe supports issues. The CAT inspection included pipe and pipe supports.

Region IV performed a series of room turnover inspections which included pipe and pipe supports. The Technical Review Team (TRT) inspected pipe and pipe supports.

At present, Stone and Webster (S&W) has full responsibility for the pipe support issues at CPSES. A whole new program is in progress and NRR has the oversight responsibility for both design and the pipe support inspections.

The third party, Energy Research Corporation (ERC) is, in addition to S&W walkdowns, performing independent inspections. The third party design organization, TERA, is also overseeing the S&W design effort. The Region IV Comanche Peak Group is performing independent inspections of the ERC effort which includes independent inspections of the pipe supports. The Office of Inspection and Enforcement has been requested by NRR to audit the S&W OA program associated with CPSES activities.

My point is that the Phillips' decision to inspect this Bulletin certainly wasn't coordinated with RIV management. I believe the pipe and pipe support effort was already being adequately covered by RIV and NRR.

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$' REolON IV 611 RYAN PLAZA DRIVE. SUITE 1000 ARLINGTON, TEX AS 79011 ,

December 2,1985 Docket Nos: 50-445/446 MEMORANDUM FOR: Gary G. Zech, Chief Vendor Program Branch ,

THRU: Thomas F. Westerman, Chief Comanche Peak Group FROM: H. S. Phillips, Senior Resident Inspector Comanche Peak

SUBJECT:

BISCO FIRE SEALS CERTIFICATION AND QA RECORDS WHICH BURNED I identified an unresolved item in NRC Report 84-22 dated October 11, 1984 which dealt with questionable certification with respect to testing of BISCO seals. Recently I asked Mr. Thomas Young to followup on this item. During the .last two weeks he has identified two statements that may be false with respect to meeting the test require-ments of ASTM E-119 and IEEE 634 as required by FSAR (Enclosure 5) and Gibbs & Hill Specification No. 2323-MS-3BF and to retaining 0A records. This work is being done under TUGC0 Contract No. CP-0707 (between TUGC0 and BISCO). This work involves the design and testing of numerous configurations of electrical penetration seals.

Gibbs & Hill specification No. 2323-MS-38F requires that the electrical penetration configuration be tested in accordance with ASTM E-119 and IEEE 634 testing requirements. While reviewing documentation, it was noted that 81500 Letter No. 3346-271 (Enclosure 1) certifies that these test requirements were met; however, a summary of test results (Enclosure 2) states that the IEEE test requirements for' Test No. PCA-76 were not considered. BISCO Seal Design Matrix (Enclosure 3) identifies Texas Utilities Services and BISCO drawings, material, and tests. It also shows approval by the American Nuclear Insurers (ANI). The NRC inspector found no objective that this design configuration was tested to IEEE 634 requirements.

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During this review, the NRC inspector learned that in October 1776 1 portions of the tests performed on Test No. PCA-76 failed and subsequent I tests demonstrated that this design configurat%n passed. In a conversation with Mr, Gary Fedor, a BISCO Development Engineer at Park Ridge, Illinois, the NRC inspector found that the ANI had rescindeo I their acceptance of Test No. PCA-76 because test data / documentation had been lost in a fire in the sumer of 1976. In BISCO letter dated September 1985 (Enclosure 4) there is a statement that it was either

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burned in a fire or lost after the fire. Since the testing was done i in October 1976, these statements appear to be conflicting or false.

The fire of 1976 was preceded by a 1975 fire.

The circumstances of these fires are interesting because (1) records were not adequately protected, (2) corrective action was not taken after the 1975 fire, (3) records of test failures were lost, (4) there is a question of how many quality records were lost, and (5) was any aspect of this reportable per 10CFR Part 21 requirements?

The ANI required BISCO to retest. Enclosure 2 shows that Test No.

748-183 (very similar to PCA-76) failed in 1985 and Test No. 748-134 (not similar; i.e. size and tray configuration) passed in 1984 The 1984 test preceded Texas Utilities request to load fuel in September 1984 and the ANI request for retest and subsequent failure was after the requested fuel load date. Two questions arise: (1) was the 1984 test appropriate for the specific configuration, and (2) does BISCO documentation indicate that Texas Utilities Equipnent Qualification (EQ) Program is adequate; i.e. are the technical reviews BISCO design and testing adequate and does their audit of BISCO verify test data to be dependable and backup test reports and certifications. If the Vendor Program Eranch perfoms an inspection, please furnish any information concerning these questions to assist in inspecting Texas Utilities EQ activities concerning BISCO.

It is recomended that the Vendor Program Branch perfom an in-depth l inspection of BISCO to determine if their QA program assures that safety related activities are controlled and penetration seals are designed and tested in accordance with requirements. It is also recommended that a determination be made as to whether certifications and statements are false and should be investigated.

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p.

3-If you have further questions regarding this matter, please contact me at (817) 897-2201, t

f H. S. Phillips. Senior Resident Inspector Comanche Peak

Enclosures:

1. BISCO Letter No. 3346-271'
2. Summary of Testing
3. BISCO Design Configuration
4. BISCO Letter Dated 9-16-85
5. FSAR cc: T. Westeman

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bisco (817) 897-4881 Ext. 801 brand Industrial services, inc.

1420 renaissance drive park ridge, Illinois 6o068 (312) 296 1200 September 16, 1985 **'*' 202'B2 Wand orid Texas Utilities Generating Co.

P.O. Box 1002 Glen Rose, Texas 76043 .

Attn: C. G. Creamer i

Subject:

Penetration Seals - TUGC0 Contract no. CP-0707 '

6entlemen:

Recently you received correspondence from American Nuclear Insurers (ANI) recinding their acceptance of Fire Testing conducted on two electrical cable tray blockouts sealed with RTV silicone foam based on designs by BISCO. These tests were conducted at Portland Cement Association in 1976. 1 The reason behind this action had to do with the large scale fire test that was conducted and because of its size requiring a separate hose stream test. Both the fire and hose stream' tests were originally

"' submitted to ANI and received their certified acceptance. Subsequent -- 9 to the test, BISCO experte.sced a fire at their facilities resulting in the moving of their offices and apparently durino this time the

__ tubject hose stream test was either destroyed or lost. This was brought ~~

_to light when ANI, unable to locate their file copy, requested a replacement from BISCO. The copy in BISCO's file was for a different

__ test and therefore total documentation for the 1976 test was not available. _

ANI requested BISCO conduct a new test to their present standards which, experienced a burn through at two hours and thirty five minutes. However, ANI has stated they will accept all fire barriers cf this BI500 design based on their prior certified acceptance, "for insurance purposes only".

AN!'s testing standards are much more . 'ingent than those required by ASTM or the Nuclear Regulatory Commision (NRC). BI500 has testing documentation that substantiates the fire seal design that failed the l ANI fire test standards, does meet and surpass the test standards of ASTM E-119 and the NRC. BISCO will supply the necessary test reports to any of their clients along with assistance in verifying their compliance with NRC fire test standards.

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