ML20054E677

Affidavit of RB Hubbard Re Breakdowns in QA Program. Significant New Info Recently Available Clearly Demonstrates Util & Major Subcontractors Failed to Develop & Implement Qa/Qc Program Complying W/Nrc Regulations

ML20054E677
Person / Time
Site:	Diablo Canyon
Issue date:	05/24/1982
From:	Hubbard R JOINT INTERVENORS - DIABLO CANYON, MHB TECHNICAL ASSOCIATES
To:
Shared Package
ML20054E672	List: ML20054E671 ML20054E677
References
ISSUANCES-OL, NUDOCS 8206140059
Download: ML20054E677 (250)

Text

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)) UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 1C H 52 2

)

In the Matter Of )

)

PACIFIC GAS AND ELECTRIC COMPANY )

) Docket Nos. 50-275 0.L.

(Diablo Canyon Nuclear Power ) 50-323 0.L.

Plant, Unit Nos. 1 and 2) )

}

AFFIDAVIT OF RICHARD B. HUBBARD CONCERNING BREAKDOWNS IN THE DIABLO CANYON QUALITY ASSURANCE PROGRAM O

8206140059 820607 PDR ADOCK 05000275 C PDR

O(_) - TABLE OF CONTENTS 1

I. INTRODUCTION . . . . . . . . . . . . . . . 1 II. BACKGROUND AND PURPOSE . . . . . . . . . . 6 III.

SUMMARY

OF CONCLUSIONS . . . . . . . . . . 9 IV. DISCUSSION OF THE SIGNIFICANT QUALITY ASSURANCE ERRORS REVEALED SINCE SEPTEMBER 21, 1981 . . .. . . . . . . . . 13 IV.A: Error 1 - Mirror Image Design Orientation . .. . . . . . . . . . 14 IV.B: Error 2 - Improper Distribution of Documents . . . . . . . . . . . 16' IV.C: Error 3 - Incorrect Weights -

Annulus Area . . . . . . . . . . . 16 IV.D: Error 4 - Containment Spray System Pipe Supports . . . . . . . . . . . 18 IV.E: Error 5 - Wrong Spectra - Piping. . 19 IV.F: Errors 6 to 10 - Additional Design Errors - Piping . . . . . . . . . . 20 IV.G: Error 11 - Incorrect Vertical Spectra - Regenerative Heat Exchanger . .. . . . . . . . . . . 22 IV.H: Error 12 - Misapplications of Hosgri Spectra - Conduit Supports . 22 IV.I: Error 13 - Further Spectra Mis-application - HVAC. . . . . . . . . 23 IV.J: Error 14 - Differences Between "As-Built" and "As-Designed" Conditions Conduit Supports . . . . . . . . . 24 IV.K: Additional Errors Disclosed Since License Suspended . . . . . . . . . 25 O

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() IV.L: Error 15 - Failure to Use Final Seismic Design Spectra -

Auxiliary Building . . .. .. . . 25 IV.M: Error 16 - Incorrect Vertical Spectra - Accumulators. . . . .. . 26 IV.N: Error 17 - Annunciator Cabinet Modeled Incorrectly . ... .. . . 27 IV.0: Error 18: Improper Digitization of Hosgri Spectra - Auxiliary Building. . . . ... .. . .. .. 27 IV.P: Error 19 - Rod IIangers Used As Vertical Seismic Restraints - Small Bore Pipes. . . .. .. . . .. . . 28 IV.Q: Error 20: Differences Between "As-Built" and "As-Designed" Conditions Piping Systems. . .. .. . . .. . 28 IV.R: Error 21 - Containment Spray Lines Drawings for Pipe Support and Valve Orientation Differ from Field Layout. . . . . . . . ... . 30 IV.S: Error 22 - Incorrect Valve Modeling in Piping Analyses. . .. . .. . . 31 IV.T: Error 23 - Raceway Supports Differ from Installation Instructions. . . 31 IV.U: Error 24 - Unconservative Electrical Seismic Criteria. . . . .. .. .. 32 IV.V: Errors Demonstrate Widespread Breakdown In Design and Construction Programs. . . . .. . . . . . .. . 33 V. IMPORTANCE OF QUALITY ASSURANCE 35 V.A: Diablo Canyon QA Review Intended to Compensate for Reduced Safety Margin . . . . .. .. . . .. . . 38 i

. _ . _ _ _ - _ _ _ _ . _ , _ _ . _ , _ . . _ . . _. . _ _ _ . _ _ _ . _ _ . _ __ . - ~ . _ _ . - _ _ . . . _ -

VI. EVOLUTION OF NRC QUALITY ASSURANCE llh REQUIREMENTS AND NUCLEAR INDUSTRY PRACTICES . . . . . . . . . . . . . . . . 43 VI.A: NRC QA Requirements Adopted in 1970 . . . . . . . . . . . . . . . 44 VI.B: Diablo Canyon QA Program Commitments Not Incorporated Into Actual Practices. . . . . . . . . . 50 VII. DEFICIENCIES IN DIABLO CANYON QA PROGRAM IMPLEMENTATION . . . . . . . . . . . . . . 5?

VII.A: QA Breakdown 1 - Failure To Establish QA Program Requirements . 55 VII.B: QA Breakdown 2 - Failure To Evaluate Suppliers. . . . . . . . . 57 VII.C: QA Breakdown 3 - Failure To Correct Conditions Adverse to Quality . . . . . . . . . . . . . . 59 VII.D: QA Breakdown 4 - Failure to Verify and Control Design Documents. . . . 61 VII.E: QA Breakdown 5 - Failure of URS/

Blume To Implement A QA Program . . 65 VII.F: QA Breakdown 6 - Failure to Conduct Adequate Design Reviews In A Timely Manner. . . . . . . . . . . . . . . 68 VII.G: QA Breakdown 7 - Failure to Control and Verify Usage of Design Data Within PGSE . . . . . . . . . . . . 72 VII.H: QA Breakdown 8 - Failure to Control Basis for m'eismic Design. . . . . . 74 VII.I: QA Breakdown 9 - QA Records Not Identifiable and Retrievable. . . . 76 l l

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VIII. PG6E AND ITS PRINCIPAL SEISMIC SAFETY-RELATED DESIGN SERVICE SUBCONTRACTORS FAILED TO ESTABLISH REQUIRED QA/QC PROGRAMS FOR WORK PERFORMED PRIOR TO JUNE 1, 1978 . . . . . ... . . . . . . . 79 VIII.A: PG6E QA Program Inadequate In Areas Of Policy, Procedures, and Implementation. . .. . . . . . 80 VIII.B: Wyle Testing Activities Lacked Controls. . . .. . . . . . . . . 82 VIII.C: Anco QA Program Not Documented. . 83 VIII.D: Harding Lawson Not Required To Implement A QA Program. . . . . . 85 VIII.E: Cygna QA Program Implementation Inadequate. . .. . . . . . . . . 86 VIII.F: Blume Failed To Establish Or Implement The Required QA Program . . . . .. . . . . . . . 87 VIII.G: EDS QA Program Implementation Adequate.-. . .. . . . . . . . . 89 VIII.H: Reedy's Conclusions Provide Significant Insights. . . . . . . 90 IX. ADEQUACY OF DIABLO CANYON QA PROGRAM AND IMPLEMENTATION FOR SITE ACTIVITIES IS UNCE RTAIN. . . . . . . .. . . . . . . . . 91 X. PGSE AND NRC PROVIDED BOARD WITH A CURSORY REVIEW OF THE EFFECTIVENESS OF THE PGGE QA PROGRAM AND IMPLEMENTATION . . 97 XI. CONCLUSION . . . . . . . . . . . . . . . . 99 iv

LIST OF ATTACHMENTS Ih ATTACHMENT DESCRIPTION A Professional Qualifications of Richard B. Hubbard B Ratio of Hosgri/DDE Accelerations Auxiliary Building C Unverified PGSE Sketch Which Led to Mirror Image Design Error D 1970 QA Program Authorization for Diablo Canyon Unit 2 E PGSE Contract Amendment of July 12, 1978 to John Blume and Associates F Initial Contract Between John Blume and Associates and PG6E Related to Diablo Canyon Dated 10/24/66 G PG6E Qualified Suppliers' List for Nuclear Safety-Related Products and Services Dated 7/78 H PGSE's Service-Related Contracts Prior to 1979 I PGSE List by Vendor of Electrical Equip-ment Requiring Environmental Qualification J PG5E Report Dated July 29, 1979 of URS/

Blume Audit K Example - Informal PGSE Internal Document Transmittal L QA for URS/Blume Work for the Diablo Canyon Proj ect M PGGE Quality Assurance Audit of Design Re-views Dated June 19, 1972 O

LIST OF ATTACHMENTS (Contd)

ATTACHMENT DESCRIPTION N PG6E Audit of Comprehensive Design Reviews Dated May 6, 1977 0 PG6E Audit of Design Verification Dated February 28, 1979 P PG6E File Review Related to Design Control c Dated October 6, 1981 Q Pipe Support and Restraint Design Guide-lines in 1976 R Design Guidelines Provided to Earthquake Engineering Systems Dated October 22, 1977 S Definition of Drawing Labels T PG6E Response to NRC Bulletin 79-14 Dated April 17, 1980 U Viewgraphs of Richard B. Hubbard for Limited Appearance Statement Before Diablo Canyon ASLB dated October 18, 1977 ,

V Limited Appearance Statement of Richard B. Hubbard Before Diablo Canyon ASLB dated October 18, 1977.

W QA Program Review Report-R.F. Reedy Audit of PG6E X QA Program Review Report-R.F. Reedy Audit of Wyle Y. QA Program Review Report-R.F. Reedy Audit of ANCO Z QA Program Review Report-R.F. Reedy Audit of Harding Lawson Associates ,

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LIST OF ATTACHMENTS (Contd) h ATTACHMENT DESCRIPTION AA QA Program Review Report-R. F. Reedy Audit of Cygna BB QA Program Review Report-R.F. Reedy Audit of Blume CC QA Program Review Report-R.F. Reedy Audit of EDS Nuclear DD PGSE Testimony Regarding QA/QC - October, 1977 ASLB Hearings vii g

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() -UNITED STATES OF AMERICA NUCLEAR REGULATORY COSEfISSION

~

BEFORE THE C0hDlISSION 4 )

In.the Matter Of )

)

PACIFIC GAS AND ELECTRIC COMPANY ) Docket Nos. 50-275 0.L.

) 50-323 0.L.

(Diablo Canyon Nuclear-Power )

Plant, Unit Nos. 1 and 2) ) ,

)

AFFIDAVIT OF RICHARD B. HUBBARD CONCERNING BREAKDOWNS IN THE DIABLO CANYON QUALITY ASSURANCE PROGRAM STATE OF CALIFORNIA )

) ss.

! COUNTY OF SANTA CLARA )

4 I. INTRODUCTION RICHARD B. HUBBARD, being of legal ~ age and duly sworn,

deposes and says as follows:

1. My name is Richard B. Hubbard. I am a Professional Quality Engineer licensed by the State of California (license i

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O number a^ 805), a technica1 consu1 tant. and a founder in 1976 and vice president of MHB Technical Associates, a corporation engaged in the business of technical consulting on energy and environmental issues, having its principal office at 1723 Hamilton Avenue, San Jose, California, 95125. I hold a B.S. in Electrical Engineering from the University of Arizona (1960) and an M.B.A. from the University of Santa Clara (1969). I have seventeen years' experience in the design, manufacture, construction, and operation of nuclear power generation facili-ties, inc1uding eleven years' experience in responsible mana-gerial positions in the Nuclear Instrumentation Department (1965-1971), Atomic Power Equipment Department (1971.1975),

and Nuclear Energy Control and Instrumentation Department (1975-1976) of the Genera 1 Electric Company (GE). For the past five years, I, along with my co-founders of hDiB Technical

' Associates, have conducted numerous studies pertaining to the safety, quality, reliability, and economic aspects of nuclear power facilities.

2. From November, 1971 to February, 1976, I was a Mana-ger of Quality Assurance for the manufacturing operations at the San Jose, California, headquarters of GE's Nuclear Energy Divi-sion. I was' responsible for the development and implementation of quality plans, programs, methods , and equipment to assure that equipment for nuclear plants manufactured and procured by O

General Electric met quality requirements as defined in NRC lll regulation 10 CFR 50, Appendix B; ASME Boiler and Pressure Vessel Code; customer contracts; and GE Corporate pclicies and procedures. The product areas include radiation sensors, reactor vessel internals, fuel handling and servicing tools, nuclear plant control and protection instrumentation systems, and control room panels for the Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP). I was responsible for approxi-mately 45 exempt personnel, 22 non-exempt personnel, and 129 hourly personnel with an expense budget of nearly 4 million dollars and an equipment investment budget of approximately 1.2 million dollars. While employed by General Electric, I was re-sponsible for developing a quality system which received NRC certification in 1975. The QA system was also successfully surveyed for ASME "N" and "NPT" symbol authorizations in 1972 and 1975, plus ASME "U" and "S" symbol authorizations in 1975.

I was also responsible for the quality assurance program and its implementation at GE's spare and renewal parts warehouse in San Jose.

3. I am a member cf the IEEE Nuclear Power Engineering standards subcommittee responsible for the preparation and revision of three Quality Assurance standards for safety-related l

aspects of nuclear power facilities as follows:

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k) a. IEEE 493 ( ANSI N4 5. 2.16) : Requirements for the Calibration and Control of Measuring and Test Equipment Used in the Construction and Maintenance of Nuclear Power Generating Stations.

b. IEEE 336 (ANSI N45.2.4): Installation, Inspection, and Testing Requirements for Class IE Instrumentation and Electric Equipment at Nuclear Power Generating Stations,

c. IEEE 467: Quality Assurance Program Require-ments for the Design and Manufacture of Class IE Instrumentation and Electric Equipment for Nuclear Power Generating Stations.

I am currently a member of the IEEE committee which is preparing a standard which addresses the requirements, including the quality assurance program, for the selection and utilization of replace-ment parts for Class IE equipment during the construction and operation phase at nuclear power stations.

4. I am familiar with the safety analyses of the Diablo Canyon Nuclear Power Plant (Diablo Canyon) prepared by the license Applicant (Pacific Gas and Electric Company), the NSSS Supplier (Westinghouse), and the Staff (NRC Staff) as a result of my service since 1976 as a technical consultant to the Center For Law In the Public Interest (CLIPI), counsel for the Joint Intervenors in the Diablo Canyon Operating License proceeding.

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As part of my responsibilities as a technical consultant for lll CLIPI, I attended the Diablo Canyon non-seismic safety issue hearings in October, 1977, including witnessing the Staff and Applicant testimony on quality assurance. In addition, I attended the majority of the sessions of the seismic safety hearings during the three month pe'riod between December, 1978 and February, 1979, and the reopened seismic hearing in October, 1980. In 1980, MHB was retained by counsel to Governor Edmund G.

Brown, Jr. to provide technical consulting services in the ongoing Diablo Canyon low power and full power license proceedings.

As a consultant to counsel for Governor Brown, I attended the low power hearings in May, 1980, and most of the full power he rings in January, 1982. I have direct personal knowledge of quality assurance breakdowns discussed herein. My k awledge related to the QA breakdowns was secured as a result of attending and participating in the PGSE/NRC meetings in Bethesda, Maryland, on October 9 and November 3, 1981, and February 3, March 25, and April 1, 1982. Personal knowledge also was obtained as a result of accompanying NRC personnel from Region V on their I inspection of the PGSE and John A. Blume and Associates (Blume) quality programs at PGSE and Blume headquarters during an eight day series of meetings in October, 1981, and as result of attending Region V meetings with Robert L. Cloud Associates, O

((() Inc. during January, 1982.* Further, I presented testimony on this matter at oversight hearings before the House Subcommittee on Energy and the Environment on November 19, 1981. In addition, I attended and made a pre.sentation at the meeting between Denton of the NRC Staff and the intervenors in the Diablo Canyon proceeding in San Francisco on February 17, 1982.

5. Finally I have testified on safety-related aspects of nuclear power facilities' quality assurance programs as an expert witness before the Nuc1 car Regulatory Commission Atomic Safety and Licensing Boards; before and at the request of the NRC's Advisory Committee on Reactor Safeguards; before the Joint Committee on Atomic Energy of the United States Congress; and before various federal and state legislative and adminis-trative bodies. A summary of my experience and qualifications is set forth in Attachment A, which is appended to this

' affidavit.

II. BACKGROUND AND PURPOSE

6. On July 17, 1981, the Atomic Safety and Licensing Board ("ASLB") issued a Partial Initial Decision ("PID")

• Certain of the October and January meetings were also attended by my colleagues at MHB Technical Associates, Mr. Dale G.

Bridenbaugh and Mr. Gregory C. Minor. Messrs. Bridenbaugh and Minor have participated in review of this affidavit.

/m concerning, inter alia, PG5E's and its maj or subcontractor's (l) compliance with the NRC's Quality Assurance regulations in 10 CFR Part 50, Appendix B. The ASLB concluded that: 1/

"the Diablo Canyon quality assurance programs for both the Design and Construction Phase and the Operations Phase have been and are in compliance with the requirements of 10 CFR 50, Appendix B, and that the implementation of both programs is acceptable to the Board."

The ASLB based its decision on evidence presented by PGSE and

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the NRC Staff at hearings on October 19, 1977.

7. Since late September, 1981, a number of serious design errors at Diablo Canyon has been disclosed. These errors reflect significant breakdowns in the Diablo Canyon Quality Assurance ("QA") and Quality Control ("QC")--/ program.

These QA/QC breakdowns led the Commission on November 19, 1981, to suspend PG5E's low power license for the facility. See Order Suspending License, CLI-81-30, November 19, 1981.

1/ PID at 11.

-*/ I proffered testimony at the October 19, 1977 hearing, but the ASLB accepted that testimony only as a limited appearance submission (PID at pp. 9 to 11). A copy of the viewgraphs and testimony I proffered in 1977 are attached hereto as Attachments U and V.

--**/ " Quality Assurance" comprises all those planned and syste-matic actions necessary to provide adequate confidence that a structure, system, or component will perform satisfactorily in service. Quality assurance includes " Quality Control,"

which comprises those quality assurance actions related to the physical characteristics of a material, structure, components, or system which provide o means to control the quality of the material, structure, component, or system to oredetermined requirements.

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() 8. At the request of counsel to Governor Brown, I have reviewed the NRC's QA/QC requirements and the factual evidence accumulated since late September, 1981 which document the QA/QC breakdowns at Diablo Canyon. The purpose of this review was to determine whether the new evidence contradicts or otherwise raises questions about the ASLB's ruling set forth in paragraph 6 on the preceding page.

9. My QA/QC review has focused chiefly on the following:

a) Examination of industry QA/QC standards prior to 1970.

b) Examination of the regulatory developments leading to the NRC's adoption in 1970 of its QA/QC regulations, c) Examination of the NRC's implementation of its QA/QC regulations in the early 1970's through issuance of Regulatory Guides, NUREG's and clarifying amendments to its regulations, and review of related industry standards, such as the ANSI Standards, d) Review of PGSE's PSAR and FSAR commitments relating to QA/QC and other evidence relating to PGSE's and its major subcontractor's implementation of a QC/QC program for Diablo i

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Canyon set forth in Region V Inspection and (l)

Enforcement Reports.

e) Review of the NRC Staff's Safety Evaluation Report and supplements thereto for Diablo Canyon dealing with QA/QC matters.

f) Review of the testimony of PGSE and the Staff at the October, 1977 hearings on QA/QC matters and the ASLB's PID relating thereto.

g) Review of the decisions of the ASLB and the Appeal Board relating to the Hosgri reanalysis and Diablo Canyon's compliance with 10 CFR Part 100, Appendix A.

h) Review of reports and investigations relating to the Three Mile Island Unit 2 accident whicn emphasize the importance of QA/QC programs in protecting the public health and safety.

III.

SUMMARY

OF CONCLUSIONS

10. My conclusions are as follows :

a) The NRC established comprehensive QA/QC regulations set forth in 10 CFR Part 50, Appendix B in 1970. These regulations, through 18 detailed criteria, and as supplemented by ANSI Standards and Regulatory Guides, specify mandatory actions which a licensee must take during design, construction, l

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() and operation to ensure that the facility, as designed and built, complies with the licensee's application and with substantive regulatory requirements. The overall requirements for quality standards and records were further emphasized in 1971 by the NRC in General Design Criterion 1 of Appendix A to 10 CFR Part 50.

b) The Part 50, Appendix B, regulations were effective immediately upon issuance in 1970. Allowing a reasonable length of time for a licensee to develop and implement a progrum in compliance with these regulations, PGSE and its major sub-contractors should have had a QA/QC program which fully satis-fied Appendix E no later than 1972.

c) The evidence documents that PGGE and its major subcontractors did not develop and implement by 1972 a QA/QC program in compliance with Part 50, Appendix B. Rather, the evidence is clear that there were major QA/QC breakdowns in 1977-78,

'and that QA/QC breakdowns continued into 1979. These breakdowns are not isolated or minor but include violations of at least thirteen of the eighteen Appendix B criteria including:

i) Criterion 1, concerning the responsibility for the establishment and execution of the QA program.

ii) Criterion 2, concerning establishing a QA program at the earliest practicable time.

iii) Criterion 3, concerning design control measures.

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iv) Criterion 4, concerning procurement document e ' trol.

g v) Criterion 5, concerning documented instruc-tions, procedures, and drawings.

vi) Criterion 6, concerning control of documents.

vii) Criterion 7, concerning conformance of pur-chased material, equipment, and services to procurement documents.

viii) Criterion 10, concerning control of inspection activities.

ix) Criterion 11, concerning control of testing activities, x) Criterion 15, concerning control of nonconform-ances to prevent their inadvertent use or in-stallation.

xi) Criterion 16, concerning the identification and correction of conditions adverse to quality.

xii) Criterion 17, concerning maintenance of records for activities affecting quality.

xiii) Criterion 18, concerning QA program audits.

d) Other, or the same, QA criteria relevant to site QA/QC mcy also have been violated. This is particularly true since site QA/QC activities were covered by the same QA manual as design QA, which has been shown to have been inadequately planned and implemented. Indeed, repeated QA program break-downs have been identified in all areas subject to the NRC's 1

narrow reinspection program, leading one to believe that site QA would also be found to be deficient if it were examined in 1

detail.

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(,) 11. PG6E's failure to establish and implement a QA/QC program in compliance with Part 50, Appendix B is a significant safety issue. First, PGSE, the Staff and the ACRS have relied upon allegedly superior QA/QC at Diablo Canyon to compensate for reduced conservatism -- reduced margins of safety -- in

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Diablo Canyon design and construction. Such reliance is misplaced since massive breakdowns in QA/QC have been disclosed.

Second, the QA/QC breakdown at Diablo Canyon means that there is no basis for confidence that safety-related structures, systems, and components have been designed and constructed in accordance with regulatory requirements and PG6E's commitments in its CP and OL applications. Accordingly, not only is there no basis for using QA/QC to compensate for less conservative licensing assumptions, the widespread breakdowns in QA/QC preclude any finding that the plant, in fact, complies with regulatory requirements. The re fore , I conclude that the ASLB's conclusion set forth in paragraph 6, above, is erroneous.

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• / The discovery of the Hosgri fault necessitated a reanalysis of Diablo Canyon design. This reanalysis led PGSE to utilize , with the Staff's support, analytical methods and assumptions which are less conservative than would be utilized on other plants. The ACRS, the Staff, and the ASLB justified use of such less conservative techniques because Diablo Canyon allegedly had been more thoroughly analyzed than most plants, making it unlikely that undetected errors would exist at Diablo Canyon. Accordingly, the alleged accuracy of Diablo Canyon design and construction -- the essence of a QA/QC program -- has been relied upon at Diablo Canyon to a greater extent than at most plants to support

,the facility's licensing.

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IV. DISCUSSION OF THE SIGNIFICANT QUALITY ASSURANCE ERRORS lll REVEALED SINCE SEPTEMBER 21, 1981

12. In the seven weeks between when the Diablo Canyon fuel loading and low power testing license was authorized on September 21, 1981, and the license was suspended on November 19, 1981 by the NRC Commissioners, at least fourteen (14) separate errors

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in design and construction were identified at Diablo Canyon.

These errors have primarily involved QA breakdowns in the devel-opment, distribution, and use of design data by PG5E and by its engineering services subcontractors. The significance of these serious QA errors cannot be disputed. Indeed, the Commission acknowledged on November 19 that "had this information been known to the Commission on or prior to September 22, 1981, Facility License No. DPR-76 would not have been issued" and accordingly the Commission suspended PGSE's license to load fuel and conduct low power tests at Diablo Canyon. Further, even when only the first two of the errors now known were revealed, Mr. Harold Denton of the NRC Staff stated that the

-*/ The precise number of errors cannot be accurately determined.

As of November 3, 1981, there were 14 errors, as acknowledged by Harold Denton in response to a question by Commissioner Ahearne. See November 9 NRC Commissioner Briefing transcript ,

p. 10. Since November 3, 1981, additional errors have been disclosed, but the number is unclear. The 14 initial errors were described by the Affiant in an attachment to Governor Brown's October 30, 1981 letter to NRC Chairman Palladino.

2/ CLI-81-30, Order Suspending License, November 19, 1981, p. 3.

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() low power license would not have been issued if these errors had been known to the Staff before the NRC issued the license. /

13. The fourteen errors were disclosed at Commission and

Staff

meetings in Washington, D.C. and Bethesda, Maryland, and at a series of meetings during October,1981, with PGSE in f

) San Francisco, California. One hundred and eightv-five (185) additional design and construction discrepancies have sub-sequently been disclosed in biweekly status reports submitted

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l by PGSE in response to CLI-81-30. As a result of these I

meetings, investigations, and reports, it is now clear that

! the errors involved a failure of PG6E and its major sub-contractors to implement properly a large number of the 18

! mandatory quality assurance criteria of 10 CFR Part 50, Appendix B. The fourteen QA errors disclosed prior to the license suspension are described below:

IV.A: Error 1 - Mirror-Image Design Orientation

14. On September 28, 1981, PG6E reported that a diagram error had been found in a portion of the seismic qualification f

3/ October 9 meeting transcript, p. 117.

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• / PGSE's thirteenth semi-monthly status report dated May 14, 1982 listed 156 discrepancies identified by the independent

design verification program and 29 additional items identified by PGGE's internal technical program. See p. 1 of " Summary" .

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of the Diablo Canyon Unit 1. This error resulted in an incor-(l) rect application of the seismic floor response spectra in the crane wall-containment shell annulus of the Unit 1 Containment Building. The error was that the diagram used to locate Vertical Seismic Floor Response (VSFR) spectra for the Unit 1 con-tainment annulus was erroneous. The diagram was applicable to Unit 2 but was identified as being that of Unit 1. Since the Units are opposite hand, this resulted in an incorrect orienta-tion of VSFR spectra for Unit 1 component and system design.

The origin of the error was in the PGSE transmittal to its principal seismic design subcontractor (John A. Blume and Associates) of an unverified, unlabeled, handwritten sketch of the Unit 2 opposite hand geometry in place of the Unit 1 geometry. $/ The improper PGSE diagram is included herein as Attachment C. Five systems were potentially impacted by the error including two ECCS systems (Safety Injection System and the Residual Heat Removal System). Other systems impacted were the Component Cooling Water Systems, Steam Generator Blowdown System, and H 2 Recombiners. E/ Blume personnel further compounded the PGSE

" sketch" error by one of its own. Blume assumed that the layout of the annulus areas of Units 1 and 2 were identical. In fact, 3/ LER 81-002/01T-0, October 12, 1981.

5/ Transcript of September 30, 1981 NRC Comm'ssioner t Briefing, pp. 21-22.

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(.l they were mirror images. $/ Affected components must now be reviewed and modified, as necessary, to comply.with the Hosgri seismic design criteria (also see Error 3).

IV.B: Error 2 - Improper Distribution of Documents

15. At the October 9 meeting between the NRC Staff and PGSE, PG6E disclosed that the Seismic Category I electrical cable trays and conduit supports had been qualified to design response spectra which had been superseded. The error was caused by PG6E's failure to distribute the latest revised spectra to the responsible engineer. The cable trays and supports were reanalyzed using the correct spectra and PGSE has reported that no modifications were found to be necessary.1/ However, certain cable trays will now be limited as to maximum permitted number of cables. 8/.

IV.C: Error 3 - Incorrect Weights - Annulus Area

16. On October 22, 1981 inspectors from the NRC's Office of Inspection and Enforcement determined that, in addition to 6/ NRC Inspection Report 50-275/81-29, p. 2.

2/ October 9 meeting transcript, pp. 105-107.

-8/ NRC Meeting Summary for October 15-16, 1981, Discussions and Preliminary Audit of Seismic Analysis for Equipment and Components in Diablo Canyon, Unit 1 Containment Annulus,

p. 2.

O the improper application of the sketch as reported to the (g)

NRC by PGSE on September 28, 1981, the weights listed on the sketch and used as an input to John A. Blume and Associates for its development of the annulus area response spectra, could not be verified as being accurate. PGSE representatives re-calculated the weights, using current as-built drawings, and determined the new weights to be different. S/ George Maneatis, Senior V.P. for PGSE, informed the NRC that: 1S!

"The original data underestimated some weights and omitted others. The revised weights are substantially heavier than those used in the development of the 1977 floor spectra." (Emphasis added).

17. PGSE concluded that the substantial weight variations resulted from three principal causes:

a) The large bore piping equipment weights were not associated with the correct frames because the Unit 1 piping orien-tation was used in conj unction with the Unit 2 frame orientation.

b) PGSE's current calculations include ad-ditional contributors to the total weight:

9/ PNO-V-81-59, Preliminary NRC Notification of Event or Unusual Occurrence, October 26, 1981.

10/ November 3 meeting transcript, p. 129.

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(/ e.g., conduit and cable trays and steel grating, which were not considered to be significant in the 1977 analysis ,

c) A more detailed calculation of large bore

) piping weights, piping support weights and equipment weights.

Since weight distribution is (..e of the crucial factors con-sidered in the determination of vertical response spectra for

! the annulus area, PG6E must now develop revised vertical spectra using the correct weight distribution data. These spectra will be compared to the spectra previously used in

! order to identify the impact upon seismic design of systems .i or equipment. 13/

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IV.D: Error 4 - Containment Spray System Pipe Supports

18. On September 18, 1981, the NRC's resident inspector was notified by telephone of a deficiency reportable under 10 CFR 50.55 (e). 12/ The followup report from PGSE addressed deficiencies in the design of the containment spray system pipe i

11/ Supplement to LER 81-002/0lX-1, October 27, 1981.

--12/ Letter from Crane of- PGSE to Engelken of NRC, October 19, j 1981. Inexplicably, this error was not brought to the NRC l Commissioners' attention on September 21, 1981 or at the-l September 30 Commissioner Briefing. In addition, the error

[ was not disclosed at the October 9 Staff /PG6E meeting. The l direct relevance of the error to the ongoing discussion of

! QA breakdowns is obvious.

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supports located within containment. The following four llh deficiencies were identified:

a) An incorrect thermal analysis was used for hanger loads.

b) This analysis was performed with one snubber modeled as a rigid member.

c) The variable spring settings for the pipe supports were improperly set, based on a deadioad analysis which assumed, incor-rectly, that the pipes contained water.

d) In designing a pipe anchor, the loads from only one side were used.

The root cause of the preceding series of errors / has not been explained by PG6E. However, it is known that as a result of the error, two anchors required structural modification and the spring supports for those two anchors required changes.

IV.E: Error 5 - Wrong Spectra - Piping

19. During the period of October 14 through 16, 1981, representatives of the NRC Staff and their consultants from

-*/ In reality, " Error 4" as described in this affidavit really consists of four distinct QA/QC breakdowns.

h I

O arooxhaven Nationa1 tadoratories met with the PGaE staff in San Francisco. During the meeting, a piping problem (PG5E #6-11) was reviewed. PG6E initially asserted that this problem did not require reevaluation as a result of the opposite hand error.

However, it was subsequently determined that the original PGGE calculation used erroneous spectra input and hence required reanalysis with the appropriate spectra. El The cause of the error has not yet been identified by PG6E.

IV.F: Errors 6 to 10 - Additional Design Errors -

Piping

20. At the November 3 meeting between PGSE and the NRC, PG6E disclosed that during its internal review undertaken as a result of the diagram frame orientation error, it had identi-fied five additional design errors requiring plant modifications

'from causes not related to the original diagram error. The design deficiencies identified are:

a) In one case, parallel piping lines which were qualified and designed from a single I analysis actually required two analyses to l

properly model both configurations.

-13/ NRC Meeting Summary for October 14-16, 1981, Discussions and

Preliminary Audit of Seismic Analysis for Equipment and Components in Diablo Canyon, Unit 1 Containment Annulus, p. 4 l

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b) In two cases, a small bore piping snubber lll required by analysis was not designed.

c) Two supports contained gaps insufficient for thermal movement.

d) A center support was not rigid in the re-strained direction.

e) A single support on a nonsafety-related pipe had not been qualified to Hosgri loads as required to prevent interaction with an adj a-cent safety-related pipe. Ad/

21. The root cause of four of the five design errors have not been determined by PGSE. The first error resulted from an error in judgement. The engineer looked at the two parallel-component cooling lines which run around the annulus, picked what he thought was the worst case, analyzed that worst case to determine pipe stressing and support loadings, and based on that the supports were designed for both lines. PG6E subse-quently determined that the engineer did not pick the worst case. Rather, the line he did not use as a model for the analysis had a longer riser section, and so it was necessary to perform a separate model for that other line. AEI 14/ November 3 mee .g transcript, pp. 138-140.

15/ November 3 meeting transcript, p. 142.

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(_/ IV.G: Error 11 - Incorrect Vertical Spectra -

Regenerative Heat Exchanger

22. A discrepancy in the spectra that were applied to the regenerative hett exchanger was discovered during Dr. Cloud's audit for PG6E. The error was that the engineer responsible for qualifying this equipment used two-thirds of the filtered spectra. 1$/

IV.H: Error 12 - Misapplications of Hosgri Spectra -

Conduit Supports

23. Error 12 involved misapplication of the Hosgri spectra.

Electrical raceway and conduit supports are unistrut type supports, are all Class 1 equipment, and are all laterally braced. The PGSE seismic analysis is based on an enveloping procedure using static analysis. In this analysis, which contains a large number of configurations, the largest weight that a particular configuration is considered to be able to have applied to it is determined, and the highest acceleration the support can experience owing to its location in the building, is also determined. Then, with those two inputs, the first 16/ November 3 meeting transcript, p. 201.

l mode frequency of the supports is calculated, and the corres- ggg ponding acceleration level is taken from the response spectra and the stress analysis is conducted. The misapplication errors were basically of two kinds. First, the analyst got the wrong number off the response spectra curve; and second, in some cases the engineer apparently used one of the Hosgri spectra from a different location in the building. 17/ -

In summary, nine of the twenty raceway support seismic calculations were found to have been done with inapplicable spe'ctra.

Further checking indicated that two of the nine may have exceeded allowable stresses if the correct spectral values were used. 18/ -

IV.I: Error 13 - Further Spectra Misapplication -

HVAC 24 For the heating, ventilating and air conditioning ~

(HVAC) system components, Dr. Cloud reviewed the seismic input for the fans and dampers. He found two instances where the Hosgri spectra was misapplied. First, calculations for supply 11/ November 3 meeting transcript, pp. 204-205.

~~18/ Preliminary Report, Seismic Reverification Program, R. L. Cloud Associates, November 12, 1981, p. 62.

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i fans S67, 68, and 69 were found to have used incorrect and unconservative seismic inputs. Second, the forced draft shutter damper qualification showed incorrect seismic definition because gravity had not been added to the vertical acceleration.

For the first error, the manner in which this analysis was conducted is very similar to that for the conduit supports (Error 12). The engineer confirmed that the equipment was rigid, and then went to the zero portion of the response spectra curve and selected the wrong value for the acceleration level. In this case, PG6E believes that the engineer used a spectra from a different location of the building. 1E/

IV.J: Error 14 - Differences Between "As-Built" and "As-Designed" Conditions - Conduit Supports

25. Several differences were discovered between the "as-built" and the "as-designed" configuration of supports for electrical conduits. The discovery of these discrepancies necessitated field verification of plant conditions as compared to the models used by PG4E for seismic analyses. 20/ -

--19/ November 3 meeting transcript, p. 206. Also, see Pre-r liminary Report, Seismic Reverification Program, p. 51.

l 20/

Progress Report No. 1, Seismic Reverification Program by R. L. Cloud, November 11, 19 81, pp . 5 and 8.

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IV.K: Additional Errors Disclosed ggg Since License Suspended

26. In the weeks since the Diablo Canyon license was suspended, numerous additional design and constructi0n defici-encies have been disclosed by PGSE and Dr. Cloud in the semi-monthly status reports to the NRC. For example, the independent design verification program has identified one hundred fifty-six (156) potential errors in the Error and Open Item (E0I) reports.

In addition, PGSE investigations conducted in response to the discovery of the initial fourteen errors disclosed twenty-nine 7

(29) further design deficiencies. a1/ As be fore , the root cause which first allowed the errors to occur, and then further allowed the errors to remain undetected for a number of years, has not been determined by PGSE. Ten (10) of the significant types of errors disclosed by PGSE and R. L. Cloud since the license was susperded are as follows:

IV.L: Error 15 - Failure to Use Final Seismic Design Spectra - Auxiliary Building

27. " Preliminary" rather than " final" seismic design spectra were utilized in the design analyses of the Auxiliary 11/ PGSE's Thirteenth Semi-monthly Status Report, May 14, 1982, p. 1 of " Summary".

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() Building and its associated equipment. The " final" design spectra curves were provided to PGSE by their consultant, URS/Blume, in October, 1979, whereas the " preliminary" design spectra curves provided to PGSE by URS/Blume in May, 1977 were apparently used in the qualification of equipment and components within the Auxiliary Building. The October, 1979 design spectra curves differ in some instances from those of May,1977. 22/

The spectra changes were largely caused by changes in the modeling of soil properties. Further, the Hosgri Amendment to the FSAR was not updated by PGSE to reflect the final design spectra. In addition, PGSE failed to promptly report to the NRC its use of superceded design data as required by 10 CFR 5 0. 5 5 (e) . The error may also apply to other Diablo Canyon structures since Blume issued approximately ten (10) final design reports for Diablo Canyon structures following the 1978/79 ASLB hearings which addressed the Hosgri reanalysis.

IV.M: Error 16 - Incorrect Vertical Spectra -

Accumulators

28. The vertical spectra input used by Westinghouse for

~~22/ NRC PNO-V-82-03, January 8, 1982. Also, see R. L, Cloud E0I 920, January 6, 1982.

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-n-- - . - - . , - - , , , . , . , . . . . - - , - - - - - , . - -

qualification of the accumulators is in error. 21/ For the ggg vertical direction, Westinghouse used two-thirds of the tau filtered horizontal spectra. The Hosgri Report at page 4-3 states that two-thirds of the unfiltered horizontal response spectra is to be used for the vertical direction. 21/

IV.N: Error 17 - Annunciator Cabinet Modeled Incorrectly

29. The original seismic analysis of the main annunciator cabinet was based on the assumption that the cabinet was rigid in the longitudinal direction. In fact, the actual first mode frequency is 14 h: or less. The result is that the seismic bracing at the top of the cabinet and the anchor bolts exceed the allowable pull out load. 2'5/

IV.0: Error 18: Improper Digitization of Hosgri Spectra -

Auxiliary Building

30. The digitization of the East-West Translational

fosgri Spectra for the 140' elevation in the Auxiliary Building has been found to contain an error. The error affected accelera-tions in the frequency range of 7.25 h: to 10.31 h: for 2%

23/ February 3 meeting transcript, pp. 123, 126 and 127, 2.

3]/ PGSE Semi-Monthly Status Report, November 25, 1981, p.

25/ R. L. Cloud E0I 949, January 20, 1982.

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(') damping and 7.58 hz to 9.90 hz for 3% damping. The affected piping analyses will be rerun and pipe supports qualified as-built or redesigned to comply with allowable piping and support stress criteria. Investigations are also in progress by PGEE to determine if other plant components are affected. 21/

IV.P: Error 19 - Rod Hangers Used As Vertical Seismic Restraints - Small Bore Pipes

31. Eleven (11) small bore pipe single rod supports were found by PG6E to be installed in locations requiring vertical seismic restraints. Since a single rod support does not restrain piping in an upward direction, it cannot qualify as a vertical seismic restraint. Modification of all Unit 1 single rod supports required to provide restraint in both vertical directions is now required. 2]/

IV.Q: Error 20: Differences Between "As-Built" and "As-Designed" Conditions - Piping Systems

52. A number of discrepancies were found during R. L.

Cloud's (RLCA) field examination to verify PG6E's as-built 26/

Letter, Crane of PG6E to Denton, et al of NRC, January 25, 1982, p. 4.

2]/ PGSE Semi-Monthly Status Report, January 8, 1982, p. 2.

drawings for piping systems as compared to the actual installed ggg configurations. These discrepancies were particularly surprising since PGSE had previously re-reviewed the piping system con-figuration in 1979/1980 in response to NRC Bulletin 79-14.

Discrepancies identified included the following:

a) The "as-built" length of line 110 from support 55S/90A to support 545/26R is shown on PGSE Design Review Isometric 446541 Revision 7 as 9-3/4 inches. RLCA field inspection showed this dimension as 12 feet. 28/

b) The length of the vertical run of line 1971 between valve 8804A and the RHR Heat Exchanger 1-1 is shown on the PGSE Design Review Isometric 446542, Revision 10, as 2 feet, 0 inches. RLCA Field inspection showed this dimension to be 2 feet, 10 inches. This item exceeds the 79-14 criteria by 2 inches. 29/

c) PGSE Design Review Isometric 446544, Revision 11, does not show the second flange on the vertical run of line 44 from the Stabill:er/ Separator.

RLCA field inspection showed the location of this second flange 13 inches above the first flange. 30/

d) Valve 8805B is shown on PGSE Design Review Iso-metric 446544, Revision 11, in a vertical position.

RLCA field inspection showed that the valve is in a horizontal position. 31/

e) The "as-built" length of line 103 south of supports 18/2R and 18/12SL is shown on PGSE Design Review Isometric 449316, Revision 3, as 15 feet, 6 inches.

RLCA field inspection showed this dimension as 6 feet. 31/

28/ R. L. Cloud E0I 933, January 20, 1982.

29/ R. L. Cloud E0I 936, January 20, 1982.

30/ R. L. Cloud E0I 937, January 20, 1982.

31/ R. L. Cloud E0I 938, January 20, 1982.

31/ R. L. Cloud E0I 940, January 20, 1982.

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() f) Support 18/7R is not shown on PGSE Design Review Isometric 449316, Revision 3. RLCA field inspec-tion showed this support to be located between valve FCV 355 and FE 46 on line 2277. 33/

g) Support 5003/V on PG6E Design Review Isometric 449316, Revision 3, is not adequately located.

RLCA field inspection showed this support to be located 18 inches north of the center line of valve FCV 4 30. 34/

IV.R: Error 21 - Containment Spray Lines Drawings for Pipe Support and Valve Orientation Differ from Field Layout

33. Two differences in the Containment Spray line have been observed between PGSE's piping design review isometric drawings and the field layout. First, while pipe support 585-23R is shown as a rigid vertical on the isometric drawing, the field verification information indicated only a dead load support.

, PG6E pipe support design criteria, Document DCM M-9, require all seismic supports to be two-way restraints with the pipe f

restrained in each direction along the restraining axis. The support will be modified by PGSE to prevent upward movement thus obtaining compliance with DCM M-9 and piping analysis requirements. 2E/ Second, Valve 9001A is ' installed in a vertical 33/ R. L. Cloud E0I 942, January 20, 1982.

34/ R. L. Cloud E0I 944, January 20, 1982.

35/ R. L. Cloud E0I 932, January 4, 1982.

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m -- ~ - - -.-~n,- , -- - - - - - - - - - , - - , , . - , - , , , _ - - ,

rather than horizontal position as shown in design review lll drawing 446540, Rev. 9. However, the deficiency does not affect piping or pipe support designs as the valve was modeled in the piping analysis with the operator in the as-built vertical position. Design Review isometric 446540 is being revised to indicate the vertical valve operator orientation. 36/ -

IV.S: Error 22 - Incorrect Valve Modeling in Piping Analyses

34. Modeling of annulus area valves was reviewed by PG5E.

Six (6) were found to be modeled incorrectly. Four (4) were modeled without including the remote operator and two were modeled with the same mass point for the remote operator support and concentrated weight, thus causing an inaccurate operator support load. As a result of these errors, the valve modeling in all Unit 1 piping analyses is now being reviewed by PGSE. 37/

IV.T: Error 23 - Raceway Supports Differ from Installation Instructions

35. Three of twenty electrical raceway supports sampled by Dr. Cloud during the as-built field verifications were found 36/ R. L. Cloud E0I 931, January 4, 1982.

37/ Letter, Crane of PGSE to Denton, et al of NRC, January 25, 1982, p. 4.

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() to differ from PG6E installation. instructions (drawings 050029 and 050030). The variances identified were as follows: 38/

a) Sample 04 - A communication problem between-design and field led to an additional 1" conduit being attached to the support.

b) Sample 15 - Incorrect bolt configuration.

c) Sample 20 - Installation in the incorrect area.

IV.U: Error 24 - Unconservative Electrical

Seismic Criteria

36. Dr. Cloud's preliminary review of the PGSE methodology.

for qualifying electrical raceway supports indicates the possibility of unconservative seismic loads in the following areas:

I a) Longitudinal support for conduit, b) Raceway stresses over the maximum span.

~

c) Effect of adjacent supports, d) Coupled frequency of the hanger and raceway.

e) Flexibility and fatigue of the joints. 22/

--38/ R. L. Cloud E0I 910, December 31, 1982. Also, see PGSE Semi-Monthly Status Report, January 25, 1982, p. 2.

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39/ R. L. Cloud E0I 930, January 4, 1982.

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IV.V: Errors Demonstrate Widespread Breakdown ggg In Design and Construction QA/QC Programs 37 The preceding series of widespread errors inescapeably leads to the conclusion that PGSE and its major subcontractors experienced serious breakdowns in their QA/QC programs during the design and construction of Diablo Canyon. Based on the preceding errors, and the additional discrepancies identified by PGSE and the independent auditors in the semi-monthly status reports, I conclude that the review to date indicates that the following general grouping of QA/QC breakdowns, and associated regulatory violations, have occurred at Diablo Canyon during design and construction:

a) Hosgri report in the FSAR does not include

- most current Blume spectra; b) Seismic spectra were not controlled by PGSE or Blume; c) Preliminary and/or incorrect spectra have been used in qualification analyses for piping and equipment; d) Electrical conduit supports installation not in compliance with the design criteria plus the design criteria may have not been properly applied; O

e) Shake-tested equipment may have lacked proper O specification concerning spectra, field location, and mounting; f) Frequency calculations for electrical panels not correct; g) Tank seismic evaluations incomplete; I h) Critical aspects of pump seismic evaluations not documented; i) Discrepancies in HVAC calculations; j) Lug stress for small bore piping in excess of allowable; k) Spacing criteria for small bore piping did not address valve bypass stations, heavy valves, and equipment nozzle loads; m) Piping analysis may result in overstress co'nditions because of layout, weights, and analytical models; n) Design control of changes to buildings not adequate, building model assumptions not docu-mented, and factors omitted from building models without adequate explanation; and o) Construction "as-built" not implemented in accordance with design documents.

In the following portion of this affidavit, the NRC's required QA/QC measures are described which, if properly applied, would l

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have either prevented these widespread errors, or would have lll prevented the errors from rem?.ining unrecognized for so many years.

V. IMPORTANCE OF QUALITY ASSURANCE

38. There can be no dispute that a linchpin in the NRC's

" defense-in-depth" approach to nuclear safety is its emphasis upon QA and QC in the design, construction, and operation of a nuclear power plant. The NRC has long recognized this fact. 30/

"The application of disciplined engineering practices and thorough management and programmatic controls to the design, fabrication, construction, and operation of nuclear power plants is essential to the protection of public health and safety and of the environment.

QA provides this necessary discipline and control.

Through a QA program that meets NRC requirements, all organizations performing work that is ultimately related to the safety of plant operation are required to conduct that work in a preplanned and documented manner; to independently verify the adequacy of completed work; to provide records that will confirm the acceptability of work and manufactured items; and to assure that all individuals involved with the work are properly trained and qualified to carry out their responsibilities."

(Emphasis added)

39. Likewise, the NRC Licensing Boards have repeatedly observed that compliance with applicable quality assurance standards is an issue of critical importance to nuclear plant sa fe ty . As the Appeal Board noted in In the Matter of Consumers Power Company (Midland Plant, Units 1 and 2) , ALAB-106, 6 AEC 182, 183 (1972), "}olne of the most significant elements of the 10/ NUREG-0774, NRC 1980 Annual Report, pp. 69-70.

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\_/ Commission's ' defense-in-depth' approach to nuclear safety is its emphasis upon quality assurance arid quality control in the construction of nuclear power plants." Another Appeal Board in In the Matter of Duke Power Company (William B. McGuire Nuclear Stations, Units 1 and 2), ALAB-128, 6 ACE 399, 410 (1973, observed that: -*/

In an area as significant as quality assurance, the record should leave no doubt as to whether the applicant is in full compliance with appli-cable criteria and, if not, the basis upon which the regulatory staff authorizes any departure from such criteria.

40. The preceding is not meant to infer that QA/QC will ensure that errors will be totally eliminated. Rather, in designing and constructing any complex facility, errors are inevitable because people are not infallible. QA and QC recognize human imperfections and thus impose a control system

. designed to detect those inevitable errors and therefore, to ensure that the facility is , in fact, designed and constructed to the highest possible standards. PG6E's witness Russel P.

Wischow made this very point at the 1977 hearings: --41/

(Wlhenyouaredealingwithpeoplewhoare imperfect, you must add to that the manage-ment control and management assurance mechanism that double-checks to see whether or not the individuals are indeed performing l */ See also, In the Matter of Consumers Power Company (Midland Plant, Units 1 and 2), LBP 74-1, 8 AEC 584, 597-600 (1974);

In the Matter of Commonwealth Edison Company (Zion Nuclear Power Plant, Units 1 and 2), LBP-73-35, 6 ACE 861, 896 (1973).

41/ Diablo Canyon ASLB, Tr. 3603.

the way they had agreed upon and in accordance wit.. the established program procedures and lll understand them.

41. The previously described errors involve a large number o f systems , components, and equipment important to the safe operation of Diablo Canyon, leading inevitably to the conclusion that there has been a serious and widespread break-down in the QA program at Diablo Canycn. My exne rience suggests that, given the succession of errors already disclosed, further investigations will almost certainly reveal more errors. The Diablo Canyon errors are particularly significant because they went undetected by PGSE and Staff inspectors for years. The NRC's regulations, particularly the eighteen QA requirments of Appendix B to 10 CFR Part 50, are specifically designed to detect such errors and thus ensure that nuclear plants are designed and constructed in accordance with all requisite requirements. For example, the NRC Staff pointed out to PGSE during the October 9, 1981 meeting that errors of the type discovered at Diablo Canyon should have been detected if the Appendix B requirements had been properly implemented by PGGE. 42/ --

As presented in Section VI and VII of this affidavit, PGSE and its subcontractors did not detect these errors because they did not conply with Appendix B. In short, PGSE and its. subcontractors violated the NRC's regulations.

42/ October 9 meeting transcript, p. 86.

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V.A: Diablo Canyon QA Review Intended to Compensate for Reduced Safety Margin

42. At Diablo Canyon the need for a QA program under Appendix B is even more important than.at other facilities.

The reason is that PGSE, the Staff, and even the ACRS have r

relied upon the alleged lack of errors in design and con-1 l struction to compensate for less conservatism in Diablo Canyon design than for other plants.

43. Procedures normally used in the design of nuclear plant structures, systems and components that provide conserva-tism in design were not. applied by the Staff in the Diablo Canyon seismic re-evaluation. In Supplement 7 of the SER, the Staff acknowledges that: $2/

ltlhe generic methods of analysis used by the

Applicant in the seismic re-evaluation as out-

lined above contain three significant relaxa-tions relative to the normal, or currently

, accepted, procedures. One relaxation is reduc-i tion of ground response spectra to account for The second building size effects. (Tau effect)

! is use of actual material strengths rather than code specified minimum material strengths. The third is allowance for ductility in structures which might be used in two specific cases and specifically justified. (Emphasis added).

Other analytical relaxations implemented by PG6E in the Hosgri re-evaluation criteria included reducing the 1.15g. " peak" 43/ Diablo Canyon SER, Supp. 7, pp. 3-22 and 3-23 instrumental acceleration to an " effective" acceleration of ggg 0.75g. 44/

In addition, in the Hosgri reanalysis a damping value of 7% was used for concrete and steel structures as opposed to the 5% damping used in the original design. $5./

Figure 43-1 sets forth such a comparison between the Hosgri reanalysis and the original DDE accelerations for the contain-ment external structure. Similar information for the Auxiliary Building is included herein in Attachment B. As a result of the foregoing analytical relaxations introduced during the Hosgri reanalysis, the horizontal acceleration values for the Hosgri reanalysis were in a number of cases lower than the corresponding acceleration values used in the original Double Design Earthquake (DDE) analysis. This is astonishing since the DDE accelerations were based on a magnitude 6.75 earthquake 12 miles away while the postulated Hosgri event is a magnitude 7.5 earthquake 3 miles from the plant. Clearly, the analytical methods used in the Hosgri reanalysis result in seismic spectra less conservative than those used in the original licensing.

44 It is also evident that the design bases and criteria utilized in the Hosgri re-evaluation of Diablo Canyon were in 44/ Diablo Canyon ASLB transcript, pp. 7014 to 7016, 45/ Diablo Canyon ASLB transcript, pp. 7185-7186.

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FIGURE 43-1 RATIO 0F HOSGRI/DDE ACCELERATIONS $$./

CONTAINMENT EXTERIOR STRUCTURE e MAXIMUM ABSOLUTE HORIZONTAL ACCELERATIONS HOSGRI DDE Elevation Ahalysis HOSGRI Analysis Ratio (g) (g) DDE (ft.)

301.64 1.99 2.08 .96 (4% reduction) 275.37 1.85 1.74 1.06 (6% increase) 258.27 1.71 1.57 1.09 (9% increase) 231.00 1.51 1.18 1.28 (23% increase) 205.58 1.30 1.36 .96 (4% reduction) 181.08 1.12 1.37 .82 (18% reduction) 155.83 .93 1.29 .72 (28% reduction) 130.58 .77 1.08 .71 (29% reduction) 109.67 .68 .79 .86 (15% reduction) e MAXIMUM ABSOLUTE VERTICAL ACCELERATIONS HOSGRI DDE*

Analysis HOSGRI Elevation Analysis Ratio (g) (g) DDE (ft.)

301.64 1.60 0.27 5.93 274.37 1.12 0.27 4.15 258.27 1.02 0.27 3.78 231.00 0.97 0.27 3.59 205.58 0.90 0.27 3.33 181.08 0.82 0.27 3.04 155.83 0.72 0.27 2.67 130.58 0.62 0.27 2.30 i 109.67 0.55 0.27 2.04 i

• No vertical dynamic analysis was made for the DDE; 2/3 the peak ground acceleration was used in design --

4 2/3 x 0.40g = 0.27g i

46/ Presentation of Dr. Henry Kuo of NRC Staff to the

! ACRS Subcommittee on Diablo Canyon, June 14, 1978.

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(])

"certain cases less conservative than those that would be used gg) for an original design." $2/ (Emphasis added) However, PGSE and the Staff have repeatedly asserted that the lack of normal conservatism in Diablo Canyon design bases and criteria was compensated for by PGSE's and the Staff's extraordinarily

+hocough analyses and attention to detail.

. For instance, the ACRS stated: $$!

"Because of the extent and depth of the Staff's revieu of the Applicant's seismic re-evaluation, the likelihood of an undetected error in the seismic analyses or design is greatly reduced."

(Emphasis added)

Similarly, the Licensing Board specifically relied upon the Staff's representation of the purported thoroughness of the Staff's review: $2/

"The Staff review of the seismic design of the DCNGS has been the most extensive we have ever undertaken. This review has extended from the basic input criteria employed through the details of myriad analyses to the implementation in final design." (Emphasis added)

The re fo re , due to the foregoing reliance on alleged accuracy at Diablo Canyon, QA is even more important at Diablo Canyon than at other nuclear facilities.

47/ ACRS Letter to NRC Chairman Hendrie, July 14, 1978, p. 3 48/ Ibid 47/.

49/

September 27, 1979 Partial Initial Decision, p. 91 (quoted NRC Staff witness Mr. Knight).

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em

(_) 45 The ASLB clearly relied upon the Diablo Canyon QA program -- particularly the purported accuracy of calcula-tions -- and the Staff's audit thereof as a basis for licensing Diablo Canyon under less conservative assumptions than normally would be applied. That licensing basis is misplaced:

a) An excentional or extraordinary level of accuracy has not been achieved. Indeed, the i level of accuracy, as demonstrated by the numerous errors detected thus far, is far less than one would consider acceptable even at a normal plant; b) PGSE and the contractors did not develop and implement a OA program in compliance with Part 50, Appendix B; and c) The Staff's audit process did not detect the multitude of shortcomines -- indeed regulatory violations -- in the implementation of PGSE's design and construction QA program, i

46 The widespread breakdowns in PGSE's QA program cannot be minimized or otherwise explained on the basis that l the NRC's QA requirements were not really in place until the l

l late 1970's. Those requirements were clearly understandable and applicable from the early 1970's, thus making it all the more disturbing that PGSE and its maj or subcontractors, even in the late 1970's , had inadequate QA controls in e ffect.

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VI. EVOLUTION OF NRC QUALITY ASSURANCE REQUIREMENTS ll)

AND NUCLEAR INDUSTRY PRACTICES 46 A nuclear power plant is a very complex installation with numerous design and construction interfaces, with each interface offering a potential opportunity for an error in communication of requirements. Thus, the QA program exists to ensure that the plant is designed and constructed in accordance with the prescribed regulations as committed by PGSE in the PSAR and FSAR, and secondly, to document this compliance. In simplicity, the QA program requires that you plan what you do, do what you plan, and document what you did.

47. QA requirements for nuclear power plant design and construction have expanded greatly since the mid-1960's when the Diablo Canyon project was initiated. Early methods of ensuring quality were largely informal, guided by a modest set of codes and standards. While prior to 1970, there was no regulatory requirement for a QA/QC program, good industry practice would have resulted in a detailed QA/QC program. For example, PG6E witness Wischow implied that PGSE operated a disciplined and controlled system (Tr. at 3599). Starting in 1970, however, a new approach was developed, requiring adherence to a great number of specific detailed NRC rules and industry standards. PGSE's vice-president Don Brand in 1979 described O

o(_) the effect of the evolving QA requirements upon the construction of Diablo Canyon as follows: SSI "We did not.... anticipate the detail in docu-mentation and independent inspection of work-manship which would be required by the NRC.

For instance, simple field changes to avoid physical interference between components (which would be made in a conventional plant in the normal course of work) had to be documented as an interference, referred to the engineer for evaluation, prepared on a drawing, approved, and then released to the field before the change could be made. Furthermore, the conflict had to be tagged, identified, and records maintained during the change process.

These change processes took time (days or weeks) and there were thousands of them.

In the interim the construction crew must move 'off of this piece of work, set up on another and then move back and set up on the original piece of work again when the nonconformance was resolved."

Thus, the QA requirements which evolved during the early 1970's were pervasive, reaching into every aspect of nuclear plant design and construction.

VI.A: NRC QA Requirements Adopted in 1970

48. On April 17, 1979, the Atomic Energy Commission (AEC) published in the Federal Register, Volume 34, No. 73, a proposed

--50/ Brand, Donald A. , PGSE testimony before the California Public Utilities Commission, Application Nos. 58911 and 58912, June 6, 1979, pp. 17-18.

gS \-l

amendment to 10 CFR Part 50 which would add an Appendix B, gg)

" Quality Assurance Criteria for Nuclear Power Plants." This Appendix was officially adopted as an NRC regulation on June 27, 1970, and published in the Federal Register, Volume 35, No. 125. Requirements of 10 CFR 50, Appendix 3, apply directly to and place responsibility on the Applicant for the establish-nent and execution of the total quality assurance program.

Appendix B provides quality assurance requirements for virtually all the activities associated with the design, construction, and operation of those structures, system, and components "important to safe ty" / from which satisfactory performance is required to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. Thus, the requirements of Appendix B apply to a wide range of interrelated activities affecting the safety-related functions of structures, systems, and components including designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying.

49 The Commission also requires each Applicant fo r a construction permit to include in its Preliminary Safety Analysis Report (PSAR) a discussion of how the applicable QA requirements l

• / Also see 10 CFR 50, Appendix A, Criterion 1. l l

l l

g l

l

e'

(_T) will be satisfied during the design, fabrication, construction, and testing of the structures, systems, and components of the facility. 51/

Likewise, the Commission requires every Applicant for an operating license for a nuclear plant include in the Final Safety Analysis Report (FSAR) a discussion of how the applicable requirements of Apprndix B will be satisfied, j2/

As will be set forth in Section VII and VIII herein, PG6E and its major subcontractors failed for a number of years to develop the required implementing procedures to assure that the SAR commitments were in fact implemented.

50 Also in 1970, the American National Standards Institute (ANSI) established committees to develop formal explicit guidance procedures for licensees and their con-tractors on how to implement the AEC QA regulations. ANSI is an umbrella organization for technical societies including the Institute of Electrical and Electronics Engineers (IEEE) and the American Society of Mechanical Engineers (ASME) whose memberships have a professional interest in and familiarity with nuclear power plants. The ANSI working groups are composed primarily of representatives of electric utilities, power plant designers, and equipment manufacturers but include i

51/ See 10 CFR 50. 34 (a) (7) .

52/ See 10 CFR 50.34(b) (6) (ii) .

Q

nominal AEC/NRC representation. Beginning in 1971 and continuing ggg through the 1970's, the ANSI committees issued over a dozen final and draft QA standards pertaining to virtually every phase of plant design and construction involving safety-related equipment. These include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, insts11ing, inspecting, and testing (see Table 50-1) . The final and some of the draft standards have been endorsed in AEC/NRC regulatory guides.

51. The majority of the implementing ANSI standards for planning, managing, and performing overall QA programs were issued and available throughout the nuclear industry in final or draft form prior to 1974 In early 1974, the AEC issued for guidance a series of QA documents which included draft or final versions / of the ANSI standards set forth in Table 50-1 (except N45.2.16 and N45.2.23) as follows:

a) WASH 1284, October 26, 1973, Guidance on Quality Assurance Requirements During the Operations Phase of Nuclear Power Plants.

b) WASH 1309, May 10, 1974, Guidance on Quality Assurance Requirements During the Construction

-*/ WASH 1283, 1284, and 1309 contained a number of draf t QA standards. As these draft standards were issued as approved American National Standards, they were endorsed by regulatory guides. The applicability of the regulatory guide versus the draft standard were addressed in the implementation section of the guide.

-4/- O

i

/~'i l U TABLE 50-1 ANSI QA STANDARDS Reg.

Standard Year Guide Year Number Adopted Sub je ct Numbe r Issued N45.2 1971 General QA requirements 1.28 1972 N45.2.1 1973 Cleaning of fluid systems 1.37 1973 N45.2.2 1972 Packaging, shipping, receiv-ing, storage and handling of equipment 1.33 1973 N45.2.3 1973 Housekeeping during con- -

struction 1.39 1973 N45.2.4 1972 Installation, inspection and tes ting of instrumentation and electric equipment 1.30 1972 N45.2.5 1974 Installation, inspection and testing of structural con-crete and steel 1.94 1975 N45.2.6 1973 Qualifications of inspection, examination and testing

. personnel 1.58 1973 N45.2.8 1975 Installation, inspection and testing of mechanical equip-ment and systems 1.116 1976 N45.2.9 1974 Collection, s torage and main-tenance of QA records 1.88 1974 N45.2.10 1973 QA terms and definitions 1.74 1974 N45.2.11 1974 QA requirements in plant design 1.64 1973 N45.2.12 1977 Requirements for auditing QA programs 1.144 1979 N45.2.13 1976 Procurement of items and services 1.123 1976 N45.2.16 1975 Calibration and control of i measuring and test None N.A.

('

\-) N45.2.23 1978 Qualification of audit personnel 1.146 1980

Phase of Nuclear Power Plants, designated the " Green" book.

O c) WASH 1283, May 24, 1974, Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants -

Revision 1, designated the " Gray" book. /

52 The NRC requested that Applicant's indicate in the Safety Analysis Report (SAR) how the guidance documents will be applied to portions of the QA program and also delineate the organizational element responsible for implementing various provisions of the respective guidance documents within each major organization in the project, including that of the Applicant ,

the architect-engineer, the nuclear steam system supplier, the constructor, and the construction manager (if other than the cons truc tor) . Alternatively, if such guidance will not be followed, the SAR was to describe specific alternative methods that will be used and the manner of implementing them and was to identify the organizations responsible for their implementation.

PGSE witness specifically testified in 1977 Lefore the ASLB that the foregoing WASH guidance and ANSI guidance referenced therein were incorporated in PGSE's QA program. E2/ As will be demonstrated hereafter, however, these guidance documents were not followed.

• / Revision 0 of the " Gray" book was issued on June 7, 1973.

~~~53/ Direct Testimony of Russell P. Wischow, pp. 6-7, referenced in ASLB transcript at 3597-98 9

e 1

/"Y l ts _j 53. Based on my experience in developing the QA program '

for nuclear plant equipment at General Electric, and my familiarity with the development of QA programs by a number of utilities, I believe that by 1972 the need for a well defined QA program including detailed implementing procedures was generally recognized. By 1974, almost all AEC guidance materials, including WASH 1254, 1309, and 1283, and related ANSI standards had been issued. Nevertheless, as late as 1977-78, such a detailed program with procedures which demonstrated how the program was to be implemented was not developed by PG6E for Diablo Canyon. PG6E's failure to develop appropriate implementing procedures in a timely fashion is directly contrary to the regulatory requirement of 10 CFR 50.34 PG6E's delay in developing the necessary Diablo Canyon QA procedures is briefly summarized in the following paragraphs.

VI.B: Diablo Canyon QA Program Commitments Not Incorporated Into Actual Practices 54 From October 14 to 23, 1981, I accompanied the NRC personnel from Region V in their special inspection to evaluate the Diablo Canyon QA programs and other management control systems in effect at PG6E during the period from 1970 to the present. While the root cause of many of PG6E's e'rrors has l

not even yet been determined, it is clear to me that a g significant contributor to the many instances of so called

" informality" discovered by the NRC was PGSE's failure to develop detailed QA procedures until the 1976 to 1978 period, b!

PGSE's QA procedures in place prior to 1978 were only general, and hence failed to take the necessary further step of describing how activities were to be accomplished. As a result, I believe the required QA activities either did not occur, or did not occur in the consistent and prescribed manner which is essential to an effective QA/QC program.

55 A provisional construction permit for Unit 1 at Diablo Canyon was issued by the AEC on April 23, 1968 with a similar permit for Unit 2 issued on December 9, 1970. On November 1, 1969, PGSE organized its Quality Engineering Section (now called QA Department) . PG6E's QA manual for Unit 2 design and construction (the " Red Book") was issued in January, 1970 The accompanying letter of authorization stated that the QA program was applicable only to Unit 2 (see Attachment D herein). However, PGSE reportedly applied the Unit 2 QA manual to Unit 1 activities to the extent possible. 5_5,/ The " extent" of implementation has not been explicitly defined by PGSE.

54/ NRC Inspection Report 50-275/81-29, see pp. 3-6 55/ Diablo Canyon FSAR, p. 17.0-1.

i T'

56. Prior to issuance of the Red Book, PGSE engineers were provided with design manuals which consisted of a summary of standard PG6E design practices and design criteria that had evolved over the years. These design guides were narrow in that they set forth in " cookbook" form guidance for the development of solutions to standard engineering problems.

The pre-1970 engineering practices did not set forth QA measures to control design information including document distribution and design verification. In addition, PGSE's Chief Civil Engineer reported that the standard PG6E design guides were not maintained up-to-date. Further, he stated that the ' . commit-ments, rather than the standard PG6E design guides, were utilized by the civil engineers in designing Diablo Canyon. 55/

Thus, it can be concluded that prior to 1970, PG6E had not implemented a design QA program either generically or specifi-cally for Diablo Canyon in accordance with eighteen measures set forth in Appendix B and particularly the design control measures described in Criterion 3 of Appendix B.

57 Based on my recent review, I believe that PCSE's QA program description in the Red Book for design and construction activities consisted of a cursory recapitalization of the 18 criteria of Appendix B. Associated PG6E implementing procedures 56/ NRC Interview with Richard Bettinger, October 15, 1981.

A similar statement was provided by James Schuyler, Vice President, Nuclear Generation.

1

were vague, and failed to provide an adequate description of lll how to accomplish the general QA measures. For example, filled out examples of standard forms were generally not included in PGSE procedures until the post 1977 time period.

58 Th e re fo re , I conclude that the PGSE QA manual and associated implementing procedures in use during the period between 1970 and 1978 did not provide the necessary guidance for design and construction activities to assure compliance with the QA measures described in 10 CFR 50, Appendix B.

Further, the QA program failed to adequately document how the applicable requirements of Appendix B will be satisfied as required by 10 CFR 50.34 VII. DEFICIENCIES IN DIABLO CANYON QA PROGRAM IMPLEMENTATION 59 During the October 1981 special inspection conducted at PGSE and Blume offices in San Francisco in which I accom-panied NRC personnel from Region V, QA deficiencies related to Errors 1 and 2 (as described in Section IV herein) with regard to the performance of the Diablo Canyon Unit 1 facility design and the implementation by PGSE of applicable criteria of Appendix B of 10 CFR Part 50 were identified by the NRC Staff as follows: 5]/

--57/ CLI-81-30, p. 2, and SECY-81-636, November 6, 1981,

p. 2 of Enclosure.

1 s%

U a) The PGSE Quality Assurance Program did not appear to effectively exercise control over the review and approval of information passed to and received from URS/Blume.

b) The PG6E Quality Assurance Program did not appear to adequately control the distribution of design information within affected PG6E design groups.

c) The PG6E Quality Assurance Program did not appear to define and implement adequate Quality Assurance procedures and controls over other service-related contracts.

60. In my opinion, the NRC conclusions are correct, but too limited. I believe that the identified breaches of QA discipline appear to be symptomatic of general flaws in the Diablo Canyon QA program. Since a key factor in assessing the potential risk resulting from operation of Diablo Canycn is the assumption of a disciplined, thorough QA program, in my opinion the inadequacies in the Diablo Canyon QA program and its implementation pose a possible significant hazard to the public health and safety and to the environment (also see Section V.A herein regarding the NRC basis for reduced safety margins for the Hosgri reanalysis of Diablo Canyon structures, systems, and components.)

61. The failure of PG6E to establish QA controls over the Blume contract and the PG6E interface therewith is extremely critical. Blume was the principal seismic contractor on the entire Diablo Canyon proj ect and most of his work was performed A)

\_ _

prior to 1978 Thus, during essentially the entire Diablo Canyon llk design period, NO QA controls were in place, despite the over-riding importance of seismic design considerations to the Diablo Canyon proj ect. The re fo re , there can be no question that the disclosure of widespread QA breakdowns is extremely significant.

Examples of QA/QC deficiencies which I identified that occurred at Blume and PGSE are summarized in the following paragraphs.

VII.A: QA Breakdown 1 - Failure to Establish QA Program Requirements

62. PGSE's major service subcontractor for the Hosgri seismic evaluation, John Blume and Associates, was not contractually obligated to any QA program requireraents until July 12, 1978 (see Attachment E) which is eight years after the 18 QA criteria of Appendix B were adopted and twelve years after Blume's first contract for engineering services on Diablo Canyon (see Attachment F). PGSE's lengthy delay in establishing QA requirements is contrary to the Criterion 4 requirement that

" procurement documents shall require contractors . . . . to provide a quality assurance program consistent with the pertinent provisions of the appendix." Further, PGSE violated the require-ment of Criterion 2 that a QA program complying with Appendix B be established "at the earliest practible time consistent with the schedule for accomplishing the activities."

() 63. Between 1966 and 1978, the agreement between PGSE and Blume was repeatedly revised via letters from PG6E. On January 8, 1979, PG6E provided Blume with a set of drawings i and compiitations for Blume's use -in its review of the seismic design of the Diablo Canyon containment structure. Following 4

disclosure of the Hosgri fault, Blume was authorized on October 16, 1974 to reevaluate the assumptions and methodology employed in the original work in light of the questions which had subsequently been raised, and especially with respect to "near field" ground acceleration. On May 21, 1976, PGSE authorized additional funds for Blume's use in determining response spectra and in applying the spectra to the plant structures which in turn determines input to equipment and components. On November 2, 1976, a letter from Blume to PG6E stated that work to be performed within the next several months would involve the containment structure. I concur with the NRC's conclusion that " contract documents prior to July 12, 1978 (between PGSE and Blume) were rather informal documents, of the letter type, and did not really address quality assurance control requirements." EE/ The informal exchanges between Blume and PG6E were in violation of the procurement document requirements of Criterion 4 that " measures shall be established 58/ NRC Inspection Report 50-275/81-29, p. 8

to assure that applicable regulatory requirements, design ggg bases, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the document for procurement of services."

VII.B: QA Breakdown 2 - Failure To Evaluate Suppliers 64 Services and equipment were provided by vendors not listed on the qualified suppliers list. PGSE's qualified suppliers' 1 sit for nuclear safety-related products and services as of July, '978 is provided herein as Attachment G. Thirty-seven companies are listed as qualified suppliers. This listing is clearly deficient. First, engineering service contractors are not included on the list although a number of such con-sultants prepared safety analyses prior to the December, 1978, through February, 1979, ASLB hearings related to seismic safety.

Engineering service contractors in addition to URS/Blume involved in seismic s1fety-related design work during this time period, and not listed on the qualified suppliers list, included EDS Nuclear, Earthquake Engineering Services (now called Cygna), ANCO (formerly Applied Nucleonics), Wyle Laboratories, Westinghouse, and Harding-Lawson.51/ A listing 59,/ Progress Report No. 2, R. L. Cloud Associates, November 24, 1981, p. 3.

g

()

g_ of all service suppliers prior to 1979, as provided by PG6E, is set forth in Attachment H. $S/

65 In addition, the qualified suppliers list appears to be deficient in that numerous suppliers of safety-related equip-ment are not included on the list. For example, PG6E's environmental qualification review produced a file by vendor for electrical equipment requiring qualification. The list prepared by PGGE is included herein as Attachment I. $1/

Suppliers of safety-related equipment not included on the PG6E qualified suppliers' list include Rosemount, Barton, General Electric, Okonite, Raychem, Conax, and Westinghouse.

66 PG6E's failure to include a number of service and equipment suppliers on the qualified suppliers list ir in violation of the Criterion 7 requirement that " measures shall be established to assure that purchased material, equipment, and services. . . . . conform to the procurement documents" and procure-ment measures "shall include provisions, as appropriate, for i source evaluation....."

60/ Progress Report No. 5, R. L. Cloud Associates, January 6, 1982, see Attachment B.

--61/ Environmental Qualification by PG6E, September, 1981, page III-7 and Appendix A4-2 and A4-3.

A V

VII.C: QA Breakdown 3 - Failure To Correct Conditions Adverse to Quality 67 Criterion 18 requires that "a comprehensive system of planned and periodic audits shall be carried out to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program." As set forth in Criterion 16, "in the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action is taken to preclude repetition." In violation of these requirements, thirteen years after URS/Blume began work on Diablo Canyon, and 9 years after Appendix B was adopted, a PGSE audit of URS/

Blume disclosed a significant breakdown in the Blume QA system, as well as in the PGSE internal QA system.

68 Enclosed herewith as Attachment J is a July 29, 1979, PGSE QA audit report of the PGSE/Blume contract covering the post-1977 period. The report documents a continuing and significant breakdown of QA at URS/Blume and at PGSE during the period following the Hosgri reanalysis. The following are excerpts from the PGSE report:

The auditors found that in 12 instances URS/

Blume was not implementing contractual and Quality Assurance Manual requirements. Due to the significant nature of the audit fTndings, it was concluded that URS/Blume had not been responsive to its commitment for a quality program. (emphasis supplied).

O

({) 69. The deficiencies which were identified covered a wide range of Blume's activities:

The deficiencies occurred in the following areas: design review, internal audits, annual reports to management, organization chart, audits of Document Control Center, computer program verification, and interface procedures.

The deficiencies also directly involved PGSE's internal QA procedures, both with respect to the Blume contract and other service contracts.

There is no clearly defined interface between URS/Blume and the Engineering Department as required by ANSI N45.2.11 (88 2.2 6 5.1). Additionally, the Engineering Department manual does not presently require that design interfaces be identified and controlled between PGandE, URS/Blume, and other architect-engineers as reauired by the Standards.

(emphasis supplied).

70. Thus, the PG6E audit found that both the Blume and PGSE QA program were deficient in significant respects.

The deficiencies identified are not minor and indicate that a problem exists in the implementation of URS/Blume's quality l assurance program. In addition to the URS/Blume deficiencies, problems were identified within the interfacing activities between PGandE and its consultants. (emphasis s upplie d) .

The foregoing also demonstrates that the serious QA problems within PGSE and with Blume were not confined to the 1977 Hosgri

()

reanalysis period, but continued at least until mid-1979 Further, in mid-1979, PGSE still was failing to interface O

properly with its service contractors.

VII.D: QA Breakdown 4 - Failure to Verify and Control Design Documents 71 Uncontrolled and unverified design data repeatedly were provided by PGSE to URS/Blume and other subcontractors in violation of Criterion 3 regarding " design control", Criterion 4 regarding " procurement document control", and Criterion 5 regarding documented " instruction, procedures, and drawings."

Further, PGSE document distribution measures did not assure, as required by Criterion 6, that documents intended for use internal and external to PGSE "are reviewed for adequacy and approved for release by authorized personnel and are distri-buted to and used at the location where the prescribed activity is performed." An example of a document which violates the preceding four QA criteria is the unverified PGSE sketch included herein as Attachment C which led to the mirror image design error.

72. My discussions with PGSE and URS/Blume personnel during the October audit by personnel from NRC Region V revealed a number of deficiencies in the manner by which PGSE developed, O

() transmitted and used design data. Examples of QA program deficiencies related to the PG6E/Blume interface include the following:

a) As set forth herein in Section VII. A, contract documents prior to July 12, 1978, were informal documents, of the letter type, and did not address quality assurance control requirements, b) PG6E did establish a Senior Civil Engineer as the principal point of contact or principal interface person with URS/Blume. However, engineers under the Senior Civil Engineer frequently communicated information to URS/

Blume without going through the Senior Civil Engineer p2,/

l c) Various types of information were communicated to URS/Blume during the conduct of meetings, i

via telephone communication, and by written i

transmittals. The written transmittals were submitted by various means including: type-written and signed letters, handwritten 62,/ NRC Inspection Report 50-275/81-29, p. 9 l C:)

transmittal forms, and by being handed ll) directly to URS/Blume personnel without the use of a covering transmittal document. $2/

d) Documents and data, specifically the " diagram" and associated weight data (see Attachment C) ,

were not well controlled and appear to have received. no design review and verification prior to submittal by PGSE to URS/Blume. $d/

e) Documents reviewed at PGSE, as well as dis-cussions with PGSE personnel, confirm that work activities on the PGSE-URS/Blume contract, as related to the development, transmittal, and use of seismic design information, and as performed by PGSE personnel, were performed in an " informal" manner. For example: the development and transmittal of certain data and information were not well controlled, certain data and information were not properly reviewed and approved prior to transmittal to URS/Blume, certain design information 63/ Ibid 61/.

64/ Ibid 61/.

O

( }) received from URS/Blume was not properly channeled to all pertinent engineers (see Attachment K herein for example), and supervisors apparently did not periodically check (audit) their design engineers to assure they had received and were using the latest design data. $5/

73 Because of PG6E's failure to control design data, I share the NRC's conclusion that: $$/

"....there is little assurance that other errors of the type identified in Section 2 of this report (incorrect application of containment annulus diagrams, inability to verify accuracy of weights listed on diagrams and used as an input for the development of seismic response spectra, and failure to use latest revision of vertical response spectra in design of conduit and cable tray supports) do not exist in work developed under the PGSE-URS/Blume Contract during the 1977-78

, time period."

74 The QA violations at the PGGE/Blume interface related to the development, transmittal and use of seismic design data 1

i were not isolated errors. In my review at PG6E with the ARC of a number of service contracts, I found, as did the NRC, that: 52/

i i 65/ Ibid 61/.

66/ NRC Inspection Report 50-275/81-29, p. 10 67/ NRC Inspection Report 50-275/81-29, p. 14 73 d

l i

".... based on the PGSE documentation examined ll) it did not appear possible to verify all the sources of information used by the contractors."

Further, I concur in part with the NRC that: 18,/

"The degree of " informality" in the review, transmittal and use of information appears to differ with each service type contractor.

However, it appears that the greatest degree of " informality" existed with the URS/Blume Contract (initiated in 1977) with lesser degrees of " informality" existing with other service contractors that performed work at least up through mid 1977 to early 1978."

However, based on the information set forth herein in Attachment J, I believe the contractors failure to implement a QA program may have still existed in mid-1979. (Also see Section VIII which describes the general failure of PG5E and its seismic design contractors to implement the required QA/QC measures.)

VII.E: QA Breakdown 5 - Failure of URS/Blume To Implement A QA Program

75. On October 21, 1981, I accompanied the NRC to the meeting at URS/Blume. David Lang, the Diablo Canyon Project Manager for Blume, served as the Blume spokesman. Listed below are my major observations:

68/ Ibid 42/, also see p. 15.

O

(m) q, a) Blume first issued its QA manual in 1976 b) Mr. Lang has memos where PGEE/Blume discussed QA requirements prior to the 1978 PGSE contract amendment. For example, Mr. Lang presented a letter dated August 30, 1977 from Mr. Lang to Mr. Schuyler of PG6E concerning the Blume QA program for the post-DDE design analyses. The letter lists QA criteria that are met and not met by Blume. (See Attachment L herein). However, Mr. Lang later stated that Blume did not follow the QA procedures outlined in the letter during 1977 Rather, Mr. Lang wrote a letter in August, 1978 applying the QA manual to future Diablo work. No attempt was made to backfit the Blume QA program to previous work completed on Diablo Canyon.

c) No QA audits were conducted by Blume prior to institution of its QA manual on the Diablo project in 1978 d) Blume personnel assumed that the annulus areas of Unit 1 and Unit 2 were identical.

They were told this by PG6E, but Mr. Lang has no records of who provided Blume this erroneous information.

l l

p/

s_

3 l

__ .. ~ _ _ _ _ . _ _ _ _ , _ __ _ _ _ _ .__ _~ ___ . _ , _ .

e) Blume personnel did not make site visits ggg to verify the seismic models as compared to the actual plant configurations.

f) PGSE did not review the Blume calculations.

Likewise, PG6E QA did not review the Blume work.

g) There is no evidence of checking by Blume of some of its design calculations, Also no signature is provided in "chk'd" blank of some of Blume calculation. sheets (elevation 114-140 for vertical accelerations) .

h) Blume had no written procedures concerning design interface with PG6E or with other subcontractors. However, all Blume trans-mittals went from Mr. Lang of Blume to Mr. Ghio of PG6E, but information came in to :nany Blume engineers from PG6E.

i) The Unit 2 drawings were more readable than Unit 1 drawings, which is why PG6E transmitted Unit 2 drawings to Blume.

j) Mr. Lang believes he received a PG6E drawing showing the incorrect fan cooler orientation.

However, he could not find the drawing or the i

I 1

l l

'; general PGSE drawings used fcr the annulus (V

area re-analysis.

k) The Blume document room does not meet QA storage requirements as listed in N45.2.9

, For example, Blume documents are stored in cardboard boxes.

1) PGSE did not provide latest version of building documents to Blume as they were revised for Blume's use in the Hosgri re-analysis.

76. Based on the preceding, I conclude that URS/Blume failed to conduct its original seismic design for Diablo Canyon, as well as the subsequent Hosgri reevaluation, in accordance with the QA measures set forth in the 18 criteria of Appendix B.

VII.F: QA Breakdown 6 - Failure to Conduct Adequate Design Reviews In A Timely Manner 77 Criterion 3 requires that " design control measures shall provide for verifying or checking the adequacy of design.. ."

Such measures "shall be established to assure that applicable regulatory requirements and design basis,....as.specified in the license application,....are correctly translated into r,r U

I specifications, drawings, procedures, and instructions." In violation of these requirements, PGSE design reviews were O either not completed, or only partially completed.

78 For example, for the containment structure design review, PGSE acknowledged that: $2/

"The QA Manual and requirements did not exist until several years into the project. Thus, it is understandable that the containment design procedures and documents did not completely agree with the QA Manual. Departures from the requirements of QA Procedures PRE-2 and PRE-6 are the absence of a design procedure document (PRE-2, Paragraph 3.2.1), design criteria memorandum (PRE-2, Paragraph 3.2.2), and not completing the design review prior to issuance of the " Approved for Construction" drawings (PRE-6, Paragraph 3.7). Preparing the described documents and retrofitting to comply with new requirements was not deemed necessary and some-times, as with the design review, was not possible."

79 PGSE internal audits conducted between 1972 and 1979 identified a continuing pattern of failure to conduct adequate design reviews in accordance with Appendix B and PGSE's own QA practices. For example, in an audit dated June 19, 1972 (enclosed herein as Attachment M) , PGSE auditors concluded in part that:

69/

E. P. Wollak, Comprehensive Design Review of the Containment Structure, February 28, 1977, see Section III entitled " General Discussion and Results."

O

() e The Audit revealed that design reviews, as required by QA Procedure No. PRE-6, are generally not being conducted by the three engineering disciplines.

e It appears to be the understanding within Civil Engineering that the Responsible Civil Engineer's review and approval of the construction drawings constitutes a design review. It is, however, question-able whether independent design reviews are conducted, since the Responsible Civil Engineers are involved in the design.

Design review reports, as required by QA Procedure No. PRE-6, have not been written by Responsible Civil Engineers.

It is there fore , impossible to verify whether the design was checked for applicable specific requirements.

e Mechanical Engineering has not conducted design reviews during the last eighteen months. They are, at the present time, avoiding the requirement in QA Procedure No. PRE-6 by issuance of " Approved for Installation" drawings instead of " Approved for Construction" drawings.

e There appears to be confusion and a lack of understanding within the Engineering Department as to what a design review constitutes and how comprehensive it should be. (emphasis added) 80 Five years later, in an audit, performed in April 1977 (see Attachment N herein), the PG6E auditors once again concluded that compliance with design review measures was inadequate in that:

e These reviews comply with the requirements of Procedure PRE-6 except that (1) some apparent design deficiencies identified by the reviews have not been resolved and (2) several design review reports lack required approvals. These activities are presently required to be complete prior to fuel loading.

f3 V

e In addition, some of the proposed resolu- llh tions to apparent design deficiencies identified by the design reviews did not appear to address the question raised by the reviewer and in several instances the resolutions were not consistent with the FSAR.

The PGSE auditors also found that design verification pro-cedures as required by ANSI Standard N45.2.11 had not yet been implemented by PG6E. PGSE acknowledged that:

. . . .the requirements for design verification are contained in ANSI N45.2.11, Section 6.

While this ANSI Standard has explicit require-ments, it would be very difficult to use directly without more specific Engineering Department implementing procedures. The Engineering Department should prepare for a transition to compliance with the Company's Quality Assurance Manual for Operating Nuclear Power Plants by preparing procedures for design verification in accordance with ANSI N45.2.11, Section 6."

81. As recently as February 1979, PGSE QA auditors identi-fied significant de ficiencies in the PG6E design verification implementation (see Attachment 0 herein) . Deficiencies identi-fied included:

e The equipment supplied under the NSSS contract is being seismically reanaly ed to the Hosgri criteria and necessary modifications are to be designed by Westinghouse under Purchase Orders 4R-10592 and 4R-10600 No quality assurance requirements were included in these purchase orders. The PGandE engineer administering these purchase orders has sent a letter requesting Westinghouse to identify any quality assurance applied to this work.

g

() e Verification of fire protection design, performed since issue of the Fire Protection Review (FSAR Amendment 51) a year and a half ago, has not been done.

e Chances to previous 1v verified design have not been verified and evaluated for effect on the overall design, e Much of the Hosgri qualification work has been contracted to consultants who are required to verify any analysis in accordance with their approved quality assurance programs. Other than controls on preparation of contracts for such t services, there are no requirements in the Engineering Manual applicable to consultant design verification.

(emphasis added) 82 The pattern is clear. PG6E failed to conduct adequate design reviews in a timely fashion as required by its own QA procedures and by Criterion 3. Also the audits conducted in accordance with Criterion 18 did not result in appropriate follow-up action to correct conditions adverse to quality as presented in Criterion 16. PG6E's own review of the PG6E design control audits, conducted on October 6 of 1981 in response to the disclosure of the mirror image error, also documents repeated audit findings related to inadequate design control measures (see pages 1 to 4 of Attachment P herein).

1 VII.G: QA Breakdown 7 - Failure To Control and Verify Usage of Design Data Within PGSE

83. Error 2 involved the improper distribution of documents O

k-s 72

within PGSE. The sequence leading up to the error was identified ggg during my participation in the NRC special inspection as follows: 1-a) "On August 3, 1977, URS/Blume submitted to PGSE, via a letter, two copies of "Diablo Canyon Containment Interior Structure -

Annulus Vertical Floor Response Spectra, 7.5m Hosgri Earthquake." This set of spectra was a revision to the previous set of spectra that had been issued in May 1977."

b) "The PGSE Senior Civil Engineer, upon receipt of the August 3 letter and new set of response spectra, added a hand-written note to the letter and sent the letter and spectra to the assigned con-tainment building engineer. The hand-written note requested the containment building engineer to get copies to involved parties."

c) " Contrary to the note type request, the letter and revised spectra were not pro-vided by the containment building engineer to the engineer responsible for the design of seismic support details for electrical conduit and cable trays in the Unit 1 and Unit 2 containment annulus areas were not based upon the final set of response spectra."

The basic cause of Error 3 appears to be the failure by PGSE to implement adequate control systems to assure proper distri-bution within PGSE of data and information received from URS/Blume, to assure all appropriate PGSE personnel received 70/ NRC Inspection Report 50-275/81-29, p. 5.

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(_) such data and information, and to periodically verify that all appropriate persons were properly using the data and information.

84 Criterion 6 requires in part that document control be established to. assure that documents, including changes, "...are distributed to and used at the location where the prescribed activity is performed." Contrary to the Criterien 6 require-ments, neither I nor the NRC could find evidence that for work related to the PG6E/Blume contract during the 1970 to 1977 time period: 21/

a) "A written procedure or document ever existed that specified how and to whom within PG6E specific information and data received from URS/Blume should be channeled after receipt by PG6E."

J b) PG6E supervisors or managers routinely or otherwise checked with (audited) their design engineers to assure they had in their possession and were using correctly the latest design input data, or c) A formalized control system ever existed that would routinely or otherwise identify for applicable design engineers the latest design input information that l had been received from URS/Blume.

VII.H: QA Breakdown 8 - Failure to Control Basis for Seismic Design 85 The PG6E design criteria for important areas of the seismic design, such as piping supports and restraints, were 71/ NRC Inspection Report 50-275/81-29, p. 5 and 6 i - . , - - _ - __.

not formalized as required by Criterion 3. Rather, the design (l) basis was provided in uncontrolled sketches and letters (see Attachment Q herein). The failure to verify and control internal PGSE design documents also resulted in incorrect design data being provided to seismic service contractors.

For example, the erroneous mirror image design information was provided by PGSE to Earthquake Engineering Systems in October, 1977 (see Attachment R herein at page R-9 for the erroneous sketch) in an uncontrolled document in a letter transmitt41.

86 Design information, as discussed in the preceding, should have been assigned drawing numbers and thus prepared, reviewed and distributed in a controlled manner as required by Criteria 3, 4, 5 and 6 Even control of " formal" drawings in the PGSE drawing system may have been lacking.

The utilization o f " Preliminary Copy" drawings , " Advance Copy" drawings, and " Expedited Copy" drawings as outlined by PGSE in September, 1971 (see Attachment S herein) provide an indication that procurement and construction activities were being conducted in advance of design completion in further violation of the Appendix B criteria.

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() VII.I: QA Breakdown 9 - QA Records Not Identifiable and Retrievable

87. The " Preliminary Report, Seismic Reverification Program" prepared for PGSE by R. L. Cloud Associates documented numerous instances where important design QA records were not identifiable and retrievable as required by Criterion 17 Examples cited by Cloud included the following:

a) "The Exterior Containment spectra was superceded by the URS/Blume report issued on June 5, 1977. However, no transmittal of this spectra could be located in the PGGE files. 72/

b) The close and informal relationship between PGSE and URS/Blume engineers resulted in sparse documentation of design information, drawings, equipment, weights, pipe load, etc. from PG6E to URS/Blume. 73/

c) Unlike the sparse transmittal documen-tation from PG6E to URS/Blume, the documentation from URS/Blume to PG6E was relatively complete. 74,/

d) PG6E's Civil Engineering file was searched for the design information transmitted from PG6E to URS/Blume on the Intake Structure l

during and prior to the Hosgri studies.

l No such information was found. 75,/

--72/ Preliminary Report, Seismic Reverification Program, November 12, 1981, p. 9.

73/_ Ibid 72/, p. 13.

74/ Ibid 72/, p. 17, 75/ Ibid 72/, p. 22 s

e) The design review for the auxiliary salt- ggg water pump compartments was dated September 1976 (log 7). It was later qualified for Hosgri as a part cf the Intake Structure.

However, no formal documentation has been found to date. 76/

f) An examination of some telecon records kept in URS/Blume's file reveals that there was controversy of weights for the DDE model computed by PGSE in the E-W and N-S directions for all elevations except at E1. 140. A difference of 35%

in the weight at Elevation 140', computed by PGSE for the DDE model and URS/Blume's computation in March 1977 was also dis-cussed. According to the PGSE responsible engineer, the weight controversy was resolved with URS/Blume. However, no documentation of resolution has been found to date. 7 7,/

g) Very little documentation (for fuel handling crane) was found in PGSE's file on design information transmitted to URS/Blume. Based upon the recollection of the PGSE responsible engineer for the seismic analysis of fuel handling crane, the latest revisions of crane manufacturer's drawings, original calculations, and material properties of crane were transmitted to URS/Blume. As is the case for some of the other structures, the information was passed on in an informal basis. 78/

h) PG5E's Civil / Structure file (for outdoor water storage tanks) was searched for the design information transmitted from PGSE to URS/

Blume. No relevant transmittals were found. . .

The design information was found to be trans-mitted informally. 79/

76/

Ibid 72/, p. 24 E/ Ibid H/, p. 32.

78/ Ibid 72/, p. 40, 79/ Ibid 72/, p. 42 O

([) i) For the scope of piping assigned to URS/Blume, very little correspondence was located during the time frame of Unit 1 piping analyses. 80/

j) No documentation has been found concerning transmittals of information (for valves) from PGSE to EES at this point in time...

Some records of EES transmittals to PG6E have been found to date. 81/

k) A limited amount of documentatien (for valves) of information transferred from PG6E to EDS has been found to date....Some records of results sent by EDS to PG6E have been located. 82/

1) Insufficient records (for valves) have been found to fully document information flow from PG6E to Westinghouse. 8 3,/

m) Hosgri duct support qualifications for the 4KV Switchgear Room HV (heating and ventilating) System have not been located as of October 28, 1981. 84/

n) No documentation has been found to date regarding formal transmittals of spectra from PG6E to Wyle Labs. 85/

o )' For equipment (electrical and instrumentation) requalified by analysis, as indicated by note 5 in Table 10-1 of the Hosgri report, no information has been found to date as to who had performed these analyses." 86/

80/ Ibid 72/, p. 46 81,/ Ibid 3/, p. 48 82/ Ibid 72/, p. 49 83/ Ibid 72/, p. 49.

84/ Ibid 72/, p. 52, 85/ Ibid 72/, p. 56.

86/ Ibid 72/, p. 58

88 Thus, PGSE has failed to comply with the requirement lll of Criterion 17 that " sufficient records shall be maintained to furnish evidence of activities affecting quality." The requirement for the maintenance of design records for the life of the plant is also required by 10 CFR 50, Appendix A, GDC 1. In summary, PGSE's record-keeping was so little or no documentation could be readily located to demonstrate that safety-related structures and components of Diablo Canyon can withstand safely the earthquake forces for which they were presumably designed.

VIII. PGSE AND ITS PRINCIPAL SEISMIC SAFETY-RELATED ,

DESIGN SERVICE SUBCONTRACTORS FAILED TO ESTABLISH REQUIRED QA/QC PROGRAMS FOR WORK PERFORMED PRIOR TO JUNE 1, 1978 89 In response to the NRC order suspending the Diablo Canyon operating license, an independent design verification program was initiated by PGSE. As one aspect of " Phase I" of the required reverification program, the consulting firm of R. F. Reedy Inc. conducted a QA program review and audit of PGSE and its principal seismic safety-related design service subcontractors who performed work prior to June 1, 1978. The resulting conclusions by Reedy have, almost without exception, O

shown a failure to institute the QA/QC program control measures committed to by PGSE in the license application and required by Appendix B to 10 CFR Part 50 The following portions of this affidavit set forth a brief summary of Reedy's conclusions and findings related to PGSE and its six principal seismic design service subcontractors.

90. In all cases, the purpose of Reedy's review and audit was to address the adequacy of the service contractor's quality assurance procedures, controls and practices concerning the development, accuracy and transmittal of seismic safety-related information by the contractor to PGSE and other consultants to PGSE. The basis of this review and audit was to determine if the contractor's QA Program, as implemented prior to June 1978, met the applicable requirements of 10 CFR 50, Appendix B, for the seismic safety-related design services performed for PGSE's Diablo Canyon Unit 1 Plant.

VIII.A: PGSE QA Program Inadequate In Areas Of Policy, Procedures, and Implementation

91. On February 23, 1982 Reedy completed the QA review

! and audit of PGSE's safety-related activities concerning the l

Diablo Canyon Nuclear Project. The results of the audit are included herein in Attachment W. The purpose of this review

()

and audit was to assess the adequacy of PGSE Quality Assurance Program prior to June, 1978 with particular emphasis on activities that could affect seismic related design. The baseline for this review and audit were the requirements of 10 CFR 50, Appendix B. The major conclusions were as follows:

a) "The PGSE Quality Assurance program for design work was not adequate in areas of policy, procedures and implementation.

The Quality Assurance organization had insufficient program responsibility.

b) A general weakness existed in internal and external interface and document controls.

This questions whether appropriate design information was being exchanged and utilized by design groups and consultants. One concern is if the latest Hosgri seismic data was inputted for design analysis, c) The design verification program was not f6rmalized and was inconsistently implemented and documented. This included maj or gaps in roach for mechanical and other equipment." design overviews of the design a 92 Specific significant programmatic and implementation deficiencies identified in PGSE's QA program included the following:

a) " Quality Assurance as defined in the QA Manual was essentially an audi.t role. The Quality Assurance group was not assigned a primary role in determining QA requirements, b) PGSE had no procedure for assuring the completeness of the QA program to address the requirements of 10 CFR 50, Appendix B.

g

'( ) c) There were no provisions for document control of correspondence and design documents, d) During Phase I (i.e., prior to June, 1978) there were no controlled procedures for design control, design interfaces and design responsi-bilities.

e) PGSE did not require design consultants to implement Quality Assurance requirements.

f) Corrective action provisions were not addressed except with respect to audit deficiencies and deficiencies at the site.

g) Indoctrination and training were not addressed in the QA Manual or procedures.

h) The QA Manual contained no provisions for PGSE management review of the QA program for status and adequacy.

93. Thus, in my opinion there can be no question that the nearly 200 design and construction deficiencies identified to date, and summarized herein in Section IV, had as a signifi-cant root cause the failure of PG6E to implemt nt the required and committed to QA/QC program. Additional serious and wide-spread breakdowns were identified in PG6E's seismic design subcontractors QA/QC programs as set forth in the following paragraphs.

VIII.B: Wyle Testing Activities Lacked Controls 94 As summarized herein in Attachment X, Reedy concluded that the PG6E contractual requirement that Wyle perform its

. s

(~J s_

seismic safety-related testing activities under the controls of a QA program was not in effect prior to December 1, 1978.

In addition, insufficient objective evidence was submitted to Reedy to indicate that Wyle's seismic testing activities performed prior to December 1, 1978 were in compliance with a QA program. Documentation of the required QA records applicable for the PGGE contracts had been destroyed without authorization during the Wyle QA Department's move from one building to another. As such, a meaningful audit of objective documented evidence of Wyle's QA program implementation was not possible.

95 Based on the preceding, the Reedy team concluded that the seismic safetv-relater! testing activities performed by Wyle for Diablo Canyon cannot be accepted as having been performed and controlled under a Quality Assurance Program which met the applicable criteria of 10 CFR 50, Appendix B.

VIII.C: Anco QA Program Not Documented

96. In the time period prior to June, 1978, Anco performed a feasibility study to determine if in-situ vibration testing could be performed on selected seismic safety-related sys tems and items to verify dynamic analysis that PGSE had performed on the systems and items. As the result of the feasibility study Anco performed, they were contracted to and did perform O

• e

(f in-situ vibration testing on selected seismic safety-related systems and items. As summarized herein in Attachment Y, Reedy concluded that a QA Program was not implemented by Anco for those seismic safety-related design activities they provided to PG6E for the pre-June 1978 time period. Specifically, Reedy found that:

a) "The Anco Quality Assurance Manual (ANCO Spec. QAM-002, May 1978) required preparation, approval and implementation of a sufficient variety and number of QA procedures whereby if accomplished by ANCO and the applicable criteria of 10 CFR 50, Appendix B, could have been achieved.

b) PG6E accepted ANCO Spec. QAM-002 in April 1978 but no documentarv evidence was made available to the surve' team that indicated PG6E made a formal request that ANCO prepare and submit to PGSE for approval the applicable Quality Assurance Implementation Procedures.

c) ANCO did not provide documentary evidence of any actions taken to review work performed prior to April 78 for compliance to the implied control requirements of ANCO Spec. QAM-002, May 1978, d) No documentary evidence was provided by ANCO to verify that activities performed from April 1978 to June 1978 were performed 3-accordance with the implied control require-ments of ANCO Spec QAM-002, May 1978. (Note:

ANCO advised that some QA files had inadver-tently been destroyed without notification to PGSE.)

Q t

VIII.D: Harding Lawson Not Required To lll Implement A QA Program 97 During the time in question, Harding Lawson Associates (HLA) performed soil investigations, geotechnical studies and consulting for the Diablo Canyon Unit 1 Plant. The principal purposes of the HLA work was to determine physical characteristics of subsurface materials, provide design criteria and specifi-cations for soil fills, provide design criteria for foundation support and to determine slope stabilities.

98 For the QA review of HLA, Reedy utilized the applicable requirements of 10 CFR 50, Appendix B and included ANSI Standard N45.2.20, " Supplementary Quality Assurance Requirements for Subsurface Investigations for Nuclear Power Plants," as the Quality Assurance criteria that were to be met. On the basis of the review, and as summarized herein in Attachment Z, Reedy concluded that:

a) "HLA was not required to implement a formal QA program for their activities prior to April 10, 1978.

b) The HLA QA program and operating procedures applicable to the activities performed prior to June 1978 did not prescribe adequate controls to comply with the applicable criteria require-ments of 10 CFR 50, Appendix B, or ANSI N45.2.20, c) There was not sufficient objective evidence .

available at HLA to establish that a controlled system was in effect which could be accepted as equivalent to 10 CFR 50, Appendix B.

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() VIII.E: Cygna QA Program Implementation Inadequate 99 Cygna (EES) performed re-analysis of selected piping systems, design review of piping supports, analyzed Class 1 piping and re-evaluated all Class 1 supports for Diablo Canyon.

The results of Reedy's review are included herein as Attachment AA. Reedy concluded that:

a) The Quality Assurance Program, EES' Quality Assurance Manuals Rev. 2 (January 30, 1976) through Rev. 4 (December 28, 1977) in comparison to 10 CFR 50, Appendix B, minimally addressed the applicable criteria.

b) Program implementation was inadequate as evidenced by the two findings and seven observations.

100. The two findings were that:

a) There was a lack of interface control regarding input from PGSE. EES could not verify that one support (out of a sample of 5) sent from PG6E had ever been analyzed or deleted as a require-ment. There was no interface control procedure.

b) " Controlled" memos were not adequately controlled as to content and distribution. Memos did not indicate whether they superseded previous memos and were not specific in referencing other material.

In addition, Reedy's observations included the following:

a) Formal records of the computer program verification of SAP IV were not available and the verification could be confirmed only with the aid of the personnel involved, b) No Quality Assurance procedures other than the Quality Assurance Manual itself were developed and implemented; for example, conducting audits.

() - -

c) Quarterly management review meetings were not held quarterly, lll d) Audits were not timely or comprehensive.

Only calculations and computer binders were audited. There was no evidence of effective corrective action.

e) Other than the basic QA indoctrination and training session there was no evidence of auditor training and qualification.

VIII.F: Blume Failed To Establish Or Implement The Required QA Program 101. During the period in question prior to June, 1978, Blume performed a number of significant tasks such as performed structural design and analysis, prepared seismic criteria, dynamic analyses of piping, reanalyses, and various consulting services for PGSE for Diablo Canyon. On the basis of its review of Blume's Manuals and documentation applicable during this time period (pre-June, 1978), Reedy concluded that:

a) Blume did not establish or implement a QA program that met the applicable requirements of 10 CFR 50, Appendix B.

b) There was no objective evidence that an equivalent program or system of controls was in effect (at Blume) during this time period.

102 The results of Reedy's review of Blume's QA controls are included herein as Attachment BB. As stated in the preceding, Reedy attempted tc' review existing objective O

() evidence against its audit checklist to determine whether or not there was documented evidence that " good engineering practices" were used at Blume which could be evaluated as an informal equivalent contra 11ed system. .The objective evidence reviewed did not support such an equivalency having been in effect prior to June, 1978, Further, there was no evidence of design inputs or design documents being controlled or of design verification being performed at Blume. In addition, on January 26, 1982, Reedy revisited Blume to determine whether or not the work performad prior to June, 1978 was reverified after that date under the provisions of an acceptable QA program. If Blume had reverified the Hosgri evaluation under the provisions of an acceptable QA program, this would provide a chance to minimize the impact of the design work performed prior to June 1978 During this visit it was determined that Blume did not do a comolete reevaluation of the Hosgri design work under this later version of the QA program. Thus, Reedy independently determined that PGSE's principal seismic subccatractor, a contractor whose personnel presented extensive testimony to the ASLB on the design adequacy of Diablo Canyon, failed to establish and implement the required QA controls.

VIII.G: LDS QA Program Implementation Adequate lll 103. The final major pre June, 1978 seismic service subcontractor to PGSE for Diablo Canyon was EDS Nuclear.

During the time in question, EDS performed a structural evaluation of pipe anchors for piping systems installed at the Diablo Canyon Unit 1 Plant. The purpose of the evaluation was to review previously designed pipe anchors for new design conditions. Reedy concluded, based on its review of the EDS Quality Assurance Manual, records and design documentation, that EDS implemented a Quality Assurance Program which met the applicable requirements of 10 CFR 50, Appendi: B and ANSI N45.2.11.

104 While Reedy found EDS's QA program and implementation were adequate, the following five observations of QA breakdowns ere also noted:

a) We could not verify the existence of the

" Design Review Criteria" used in accordance with QAP 3.7, paragraph 3.1 for the July 26, 1978 Design Review. Evidence exists that this procedure was followed, but the docu-mentation from 1977 was not saved. There fore it is not possible to directly verify that the

" Design Review Criteria" document met all the EDS requirements for the July 26, 1978 Design Review.

b) In many cases, " point of use" references were not made for equations in the calculation files.

These references were required by QAP 3.4, paragraph 3.2.1 of the present EDS Quality Assurance Manual.

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(_s) c) We could not verify that the Quality Assurance Manager concurred with the resolution of discrepancies as required by QAP 3.7, paragraph 3.5, because there were no sign-offs on the documentation.

d) We were unable to verify that project personnel had read the project interface instructions as required, but did not find any interface problems of significance.

e) Memoranda were used in lieu of technical instructions. This is not consistent with the EDS Quality Assurance Program.

The summary of Reedy's review of EDS is included herein as Attachment CC.

VIII.H: Reedy's Conclusions Provide Significant Insights 105. The multiplicity of QA/QC breakdowns disclosed in Reedy's limited QA/QC audit of PGSE and its major seismic design con actors provide clear insights into "why" the nearly 200 design and construction discrepancies occurred and remain undetected. Clearly, the required management controls of QA/QC were either missing or not fully implemented.

The commitments to QA/QC in the Diablo Canyon license application were shown to be empty promises.

106 The broad QA/QC program requirement of 10 CFR 50, Appendix B were adopted in 1970 and have remained essentially unchanged since that time. Thus, the Reedy review of the Diablo Canyon QA/QC program against these broad regulatory l

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(

criteria, rather than a comparison of implementing procedures ll) to the more detailed guidance provided in the ANSI Standards and Regulatory Guides, was a prudent and reasonable decision.

Due to the absence of the broad QA/QC controls required by the regulations, little further insight would have been provided by such a detailed review of the implementing pro-cedures. In my opinion, the significance of the QA deficiencies disclosed to date at Diablo Canyon should not be characterized as a failure in the dotting of "i's" and crossing of "t's", but rather the significance was the disclosure of the nearly complete void of required QA/QC management controls during design activities by PCSE and its major subcontractors over almost a 10 year period. In short, the Diablo Canyon QA/QC de ficiencies were widespread and pervasive, rather than deficiencies requiring narrow fine tuning of a fully functioning QA/0C program.

IX. ADEQUACY OF DIABLO CANYON QA PROGRAM AND IMPLEMENTATION FOR SITE ACTIVITIES IS UNCERTAIN 107 There is also substantial uncertainty surrounding the actual quality level achieved at the site during the construction of Diablo Canyon. The following portion of this a f fidavit provides a brief summary of some of the significant O

breakdowns in the Diablo Canyon QA program implementation which have occurred at the site.

108 At the outset, it should be recognized that the NRC has assumed that "the weaknesses in the Diablo Canyon QA program are focused in the service contract area and that it is not a universal breakdown through the Company." 82/ I disagree with the NRC assumption. There is insufficient data to support a conclusion that QA/QC for site activities, particularly construction, met Appendix B requirements. Indeed, repeated QA program breakdowns have been found in all areas subject to the NRC's narrow reinspection program, leading one .

to believe that site QA would also be found to be deficient it it were examined in detail. Thus, there is no evidence that site QA may not have experienced the same basic problems as design QA. This is particularly true since site QA/QC activities were covered by the same QA manual as design QA, which has been shown to have been inadequately planned and implemented.

109 Criteria 10 and 11 require in part that inspections and tests be conducted to verify conformance with drawings.

Contrary to this requirement, there have been numerous instances of differences between the "as built" and "as designed" con-figuration of the plant. For example, the seismic re-analysis 87/ November 9 NRC Commission meeting transcript, p. 22.

(]}

of the Diablo Canyon piping and related pipe supports was the ggg subject of testimony before the ACRS and the ASLB in 1978 and 1979 However, following the hearings, PGSE conducted an inspection at Diablo Canyon "to verify that actual configurations of safety-related piping agree with the models used to seismically analyze them" as required by NRC Bulletin 79-14 The results of the PGSE review are shocking (See Attachment T). PG5E concluded in part that:

"The following types of discrepancies are typical of those found, in order of frequency of occur-rence: valve weights not correct; weights of valve flanges not modelled; center of gravity of valve operator not adequately considered; support location differences of greater than one pipe diameter; supports missing or extra; presence of high density lead foam or grout in penetrations; differences in pipe geometry; invalid assumptions in modeling of analysis endpoints; differences in insulating thickness and pipe diameter. It was decided that 49 of the 192 large diameter analyses and 8 of the 30 small diameter analyses had differences significant enough that the results were not obviously conservative and that they should be reanalyzed.

This amounted to approximately a 26 percent reanalysis rate. In addition, there were 10 large diameter and 4 small diameter analyses for which differences were resolved by a field hardware change." (Emphasis added) 110. Thus, in the preceding example, contrary to the requirements of Criterion 15 of Appendix B, PGSE failed to establish measures to control materials, parts, or components which do not conform to requirements to prevent their inad-vertent use or installation during construction activities at Diablo Canyon.

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111. Dr. Cloud, as discussed herein in Section IV, has

() found a number of discrepancies between the "as built" and "as designed" plant configuration of important items including, piping supports, electrical cable raceway supports and the Containment Spray line layout. The failure of PG5E and its site contractors to assure that the plant was constructed in accordance with the drawings indicates a clear deficiency in construction QA or, at least, in the interface between design and construction QA. The QA breakdown is further magnified since the pipe supports and anchors were supposedly nearly 100%

reinspected in response to the deficiencies discovered by the NRC ISE during their audits at Diablo Canyon between 1976 and 1978 112. There have been numerous additional past instances of construction defects at Diablo Canyon and allegations of construction QA problems. Examples included inadequate training of welders and radiographers, widespread defects in welding of pipe supports and pipe whip restraints, and over 10,000 defective anchor bolt installations. Of equal sig-nificance, is the fact that these QA deficiencies were not discovered (by PGSE or the NRC) until the construction of the plants was essentially complete. -/

• / For further details, see Attachments U and V to this affidavit.

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113. Another independent measure of the effectiveness of (g) the implementation of the Diablo Canyon QA program was provided by the General Accouting Office (GAO). GA0 found that five of the seven items reviewed at Diablo Canyon were deficient. $$/

The GAO audit was intended to provide an independent assessment of the effectiveness of the NRC inspection program in assuring that PGSE's QA program commitments were implemented in a thorough, disciplined manner.

114 The GAO audit occurred during the weeks of May 30 and Jane 6, 1977 As part of the audit, GA0 conducted interviews with numerous craftsmen engaged in safety-related work. The GA0 on-site audit identified safety concerns regarding (a) pipeway welding quality and structural steel installation; (b) seismic and pipe whip restraint installation; (c) adequacy of concrete anchor bolt testing; and (d) resin filter trap welding quality.

During the subsequent NRC Staff investigation, some of the allegations were substantiated, although the majority of the allegations were found by the NRC to have been previously addressed in a satisfactory manner by PG6E or the NRC. However, the NRC did cite PGSE for a regulatory violation for failure to control pipe rupture restraint documents.

115. The failure of PGSE to provide documented instruc-tions, a failure repeatedly identified in Sections VII and VIII 38/ GAO Report EMD-78-80, dated September 7, 1978, p. 13 O

of this affidavit, recently occurred during site activities.

)

As a result of the NRC inspection of October 2, through October 30, 1981, the following violation was identified: 8p/

" Facility Operating License No. DPR-76, Section 2.C.(2), incorporates the Technical Specifications into the license. Technical Specification 6.8 states that: " Written procedures shall be established, implemented and maintained covering...

applicable procedures recommended in Appendix "A" of Regulatory Guide 1. 33, Revision 2, February 1978...Each procedure...and changes thereto, shall be reviewed by PSRC and approved by the Plant Manager prior to implementation and reviewed periodically as set forth in administrative procedures." Further, " Maintenance that can affect the performance of safety-related equipment should be properly preplanned..." per section 9.A of Appendix "A" to Regulatory Guide 1.33. Appendix 4 of Section 1.A.1 to Nuclear Plant Administrative Procedure E-4 Revision 3, details that: " Written procedures shall be provided for removal or replace-ment of faulty equipment having nuclear safety significance when the plant or system is in operation. That is, removal or replacement of equipment which, if incorrectly accomplished, could... activate engineered safeguards,...or affe ct the proper functioning of engineered safeguards features be performed in accordance with written procedures." This procedure also deals with the use of sub-tier procedures, such as, " shop followers," and in Section 6 states that: " Shop followers shall not be used in lieu of a detailed written procedure...when the job complexity is such that an approved detailed procedure is warranted."

116. Thus, contrary to the QA requirements of Criterion 5 of Appendix B, there was no adequately reviewed and approved 89/ NRC Inspection Report 50-275/81-27, Appendix A.

p)

L.

procedure for repair and replacement work on the solid state lll protection system. As a result, an inadvertant safety injection signal was experienced on October 5, 1981 The root cause which contributed to the human error was the use of a shop follower in lieu of a detailed, written, PSRC reviewed, plant manager approved procedure.

117 Since there can be little or no confidence in the quality of the " paper" documenting the Diablo Canyon design, and the quality of the plant's construction implementation is likewise uncertain, it is imperative that the actual quality of all installed safety-related structures, systems, and components be established. The means to do so would be through a comprehensive design verification and physical reinspection of safety-related electrical and mechanical systems. Guidelines for the initial phase of such an independent physical reinspection and design verification audit program for Diablo Canyon were attached to Governor Brown's October 30 letter.

X. PGSE AND NRC PROVIDED BOARD WITH A CURSORY REVIEW OF THE EFFECTIVENESS OF TIIE PGSE QA PROGRAM AND IMPLEMENTATION 118 During the 1977 Board hearings, PG6E and the NRC provided testimony on the Diablo Canyon quality assurance O

(]) program in response to a Board Order. Testimony by Intervenors was not allowed by the Board, and previous Intervenors' efforts to introduce a QA contention had been denied by the Board.

Further, as a result of the Board's denial of the contention no discovery related to QA was allowed, and consequently, no basis could be developed to enable informed cross-examination of PG6E and NRC witnesses.

119 A copy of PGSE's seven pages of QA testimony which was presented to the Board in October, 1977 is attached hereto as Attachment DD. Today it is clear that the cursory seven page description of the Diablo Canyon QA program provided in PGSE's 1977 testimony for design, construction, and operation QA activities was in error, and potentially misleading, in that the testimony failed to reflect the actual status of the QA program and its implementation. For example, PG6E witness Mr. Wischow stated:

"The program for design, construction, and startup of all Design Class I structures, systems and components requires that all Company Departments ,

contractors and suppliers establish and maintain in effect quality assurance programs appropriate to the importance of their activities to safety.

Requirements for contractors' and suppliers' quality assurance programs are prescribed in design specifications. Specified requirements are based on Appendix B, 10 CFR 50 Contractors and suppliers are not permitted to proceed with their work until they have submitted a Quality Assurance Manual describing their quality assurance program and received approval from PGandE." (ps 6)

We now know that this statement was simoly not true. The lll ASLB's reliance on PGSE's, and the NRC Staff's, mistaken testimony and assurances, resulted in a seriously flawed decision.

XI. CONCLUSION 120. Based on the foregoing, I conclude that the sig-nificant new information set forth herein which has recently become available clearly demonstrates that PGSE and its major subcontractors failed to develop and implement a QA/QC program during the design and construction of Diablo Canyon which complied with the NRC's regulatory requirements. Indeed, nearly 200 examples have been set forth herein which document PGSE's failure to provide a QA/QC program for design and site activities in a timely fashion in compliance with the license application and the regulations for activities conducted prior and subsequent to the 1977 Board hearings. Further, the result of the mistaken assurances concerning the comprehensive-ness of the Diablo Canyon QA program from PGSE and the NRC Staff is that the Board issued a seriously flawed decision.

The magnitude of significant design and construction discre-pancies disclosed to date, and the widespread serious break-down in management of the QA/QC program by PGSE and its major

() subcontractors, documented herein, vividly illustrate the substantial uncertainty in the actual quality level achieved in design, construction, and installation of all safety-related structures, systems, and components at Diablo Canyon. A complete design verification and physical inspection of all Diablo Canyon safety-related structures, systems, components, and other important safety features is now both necessary and i

prudent. The results of such a design review and site inspection should be subject to the scrutiny of the Board and all parties in the ongoing Diablo Canyon licensing proceeding.

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l 121. Further the deponent sayeth not, h ea4/9 fad 6 I Richard B. Hubbard Subscribed and sworn to before me this M ' day of Ong I

, 1982.

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() LIST OF ATTACHMENTS-ATTACHMENT DESCRIPTION A Professional Qualifications of Richard B. Hubbard B Ratio of Hosgri/DDE Accelerations Auxiliary Building C Unverified PGSE Sketch Which Led to Mirror Image Design Error D 1970 QA Program Authorization for-Diablo Canyon Unit 2

! E FGSE Contract Amendment of July 12, 1978 to John Blume and Associates i F Initial Contract Between John Blume and Associates and PGSE Related to Diablo l

Canyon Dated 10/24/66

! G PGSE Qualified Suppliers' List for Nuclear Safety-Related Products and Services Dated 7/78 H PG6E's Service-Related Contracts Prior to 1979 I PGSE List by Vendor of Electrical Equip-ment Requiring' Environmental Qualification J PGSE Report Dated July 29, 1979 of URS/

Blume. Audit i K Example - Informal PGSE Internal Document Transmittal L QA for URS/Blume Work for the Diablo Canyon Proj ect M PGSE Quality Assurance Audit of Design Re-views Dated June 19, 1972

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LJ LIST OF ATTACHMENTS (Contd)

ATTACHMENT DESCRIPTION N PG6E Audit of Comprehensive Design Reviews Dated May 6, 1977 0 PGSE Audit of Design Verification Dated j February 28, 1979 P PG6E File Review Related to Design Control Dated October 6, 1981 Q Pipe Support and Restraint Design Guide-lines in 1976 R Design Guidelines Provided to Earthquake Engineering Systems Dated October 22, 1977 S Definition of Drawing Labels T PGSE Response to NRC Bulletin 79-14 Dated April 17, 1980 U Viewgraphs of Richard B. Hubbard for Limited I,7pearance Statement Before Diablo Canyon ASLB dated October 18, 1977 V Limited Appearance Statement of Richard B. Hubbard Before Diablo Canyon ASLB dated October 18, 1977.

W QA Program Review Report-R.F. Reedy Audit of PGSE

, X QA Program Review Report-R.F. Reedy Audit l

of Wyle Y QA Program Review Report-R.F. Reedy Audit o f ANCO

Z QA Program Review Report-R.F. Reedy Audit of Harding Lawson Associates

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LIST OF ATTACHMENTS (Contd) g ATTACHMENT DESCRIPTION AA QA Program Review Report-R. F. Reedy Audit of Cygna BB QA Program Review Report-R.F. Reedy Audit of Blure CC QA Program Review Report-R.F. Reedy Audit of EDS Nuclear DD PG6E Testimony Regarding QA/QC - October, 1977 ASLB Hearings

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l O

1 ATTACHMENT A PROFESSIONAL QUALIFICATIONS O F_

RICHARD B. HUBBARD i

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e PROFESSIONAL QUALIFICATIONS OF RI CH ARD B . HUBBARD O

RICHARD B. H UBB ARD MHB Technical Associates 1723 Hamilton Avenue Suite K San Jose, California 95125 (408) 266-2716 E XP E RI EN CE :

9/76 - PRESENT l Vice-President - MHB Technical Associates, San Jose, California.

l Founder, and Vice-President of technical consulting firm. Special-is ts in independent energy assessments for government agencies, particularly technical and economic evaluation of nuclear power facilities. Consultant in th is capacity to Oklahoma and Illinois Attorney Generals, Minnesota Pollution Control Agency, German Ministry for Research and Technology, Governor of Colorado, Swedish Energy Commission, Swedish Nuclear Inspectorate, and the U.S.

Department of Energy. Also provided studies and testimony for various public interest groups including the Center for Law in the Public Interest, Los Angeles; Pu blic Law Utility Group, Baton Rouge, Louisiana; Friends of the Earth (F0E), Italy; and the Union of Concerned S cien tis ts , Cambridge, Massachusetts.

Provided testimony to the U.S. Senate / House Joint Committee on Atemic Energy, the U.S. House Committee on Interior and Insular Affairs, the California Assembly, Land Use, and Energy Committee, the Advisory Committee on Reactor Safeguards, and the Atomic S af ety and Licensing Board. Performed comprehensive risk analysis of the accident probabilities and consequences at the Barseback Nuclear Plant for the Swedish Energy Commission and edited, as well as contributed to, the Union of Conce rned S cientis t's technical l

review of the NRC's Reacto r S af ety S tudy (WASH-1400).

2/76 - 9/76 Consultant, Project Survival, Palo Alto, California.

i Volunteer work on Nuclear Safeguards Initiative campaigns in Cali-

! fornia, Oregon, Washington, Arizona, and Colorado. Numerous presentations on nuclear power and alternative energy options to c ivic , government, and college groups. Also resource person for public service presentations on radio and television.

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5/75 - 1/76 Manager - Quality Assurance Section, Nuclear Energy Control and O Instrumentation Department, General Electric Company, San Jose, California.

Report to the Department General Manager. Develop and implement quality plans, programs, methods, and equipment which assure that products produced by the Depart =ent meet quality requiremen ts as defined in NRC regulation 10 CFR 50, Appendix B, ASME Boiler and Pressure %ssel Code, customer contracts, and GE Corporate policies and procedures. Product areas include radiation sensors, reactor vessel internals, fuel handling and servicing tools, nuclear plant control and pro tection instrumentation sys tems , and nuclear steam supply and Balance of Plant control room panels. Responsible for approximately 45 exempt personnel, 22 non-exempt personnel, and 129 hourly personnel with an expense budget of nearly 4 million dollars and equipment investment budget of approximately 1.2 million dollars.

11/71 - 5/75 Manager - Quality Assurance Subsection, Manufacturing Section of Atomic Power Equipmen t Departmant, General Electric Company, San Jose, California.

Report to the Manager of Manufacturing. Same functional and product responsibilities as in Engagement #1, except at a lower organizational report level. Developed a quality system which received NRC certification in 1975. The system was also success-fully surveyed for ASME "N" and "NPT" symbol authorization in 1972 and 1975, plus ASME "U" and "S" symbol authorizations in 1975.

Responsible for from 23 to 39 exempt personnel, 7 to 14 non-exempt personnel, and 53 to 97 hourly personnel.

3/70 - 11/71 Manager - Application Engineering Subsection, Nuclear Instrumen-tation Department, General Electric Co m p any , San Jose, California.

Responsible for the post order technical interface with architect engineers and power plant owners to define and schedule the instru-mentation and control sys tems for the Nuclear S team Supply and Balance of Plant portion of nuclear power generating stations.

Responsibilities included preparation of the plant instrument list with appro ximate location, review of interface drawings to define functional design requirements, and release of functional require-ments for detailed equipment designs. Personnel supervised included 17 engineers and 5 non-exempt personnel.

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12/69 - 3/70 s - -

( Chairman - Equipment Room Task Force, Nucle ar Ins trumen tation Department, General Electric Company, San Jose, California.

Responsible for a special task force reporting to the Department General Manager to define methods to improve the quality and reduce the installation time and cost of nuclear power plant control rooms. Study resulted in the conception of a factory-fabricated control room consisting of signal conditioning and operator control panels mounted on modular floor sections which are completely assembled in the factory and thoroughly tested for proper operation of interacting devices. Personnel supervised included 10 exempt personnel.

12/65 - 12/69 Manager - Proposal Engineering Subsection, Nuclear Instrumentation Department, General Electric Company, San Jose, California.

Responsible for the application of instrumentation systems for nuclear power reactors during the proposal and pre-order period.

Responsible for technical review of bid specifications, preparation of technical bid clarifications and exceptions, definition of material list fo r cost estimating, and the "as sold" review of contracts prior to turnover to Application Engineering. Personnel supervised varied from 2 to 9 engineers.

8/64 - 12/65 Sales Engineer, Nuclear Electronics B us ine s s Section of Atomic Power Equipment Department, General Electric Company, San Jose, California.

Responsible for the bid review, contract negotiation, and sale of ins trumentation sys tems and components for nuclear power plants, test reactors, and radiation hot cells. Als o ~ responsible for industrial sales of radiation sensing systems for measurement of chemicat properties, level, and density.

10/61 -

8/64 Application Engineer, Low Voltage Switchgear Department, General Electric Company, Philadelphia, Pennsylvania.

Responsible for the application and design of advanced diode and silicon-controlled rectifier constant voltage DC power systems and variable voltage DC power systems for indus trial applications .

Designed, followed manufacturing and personally tested an advanced SCR power supply for product introduction at the Iron and S teel Show'.

Project Engineer for a DC power system for an aluminum pot line sold to Anaconda beginning at the 161KV switchyard and encompassing all the equipment to convert the power to 700 volts DC at 160,000 amperes.

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9/60 - 10/61 CE Rotational Training Program g Four 3-month assignments on the GE Rotational Training Program for college technical graduates as follous:

a. Installation and Service Eng. - Detroit, Michigan.

Installation and startup t e s *. in g of the world's largest automated hot strip steel mill.

b. Tester - Industry Control - Roanoke, Virginia.

Factory testing of control panels for control of steel, paper, pulp, and utility mills and power plants.

c. Engineer - Light Military Electronics - Johnson City, New York.

Design of ground support equipment for testing the auto pilo ts on the F-105.

d. Sales Engineer - Morrison, Illinois.

S ale of appliance controls including range timers and refrigerator cold controls.

E D UC AT I ON :

Bachelor of Science Electrical Engineering, University of Arizona, 1960.

Master o f Business Administration, University of Santa Clara, 1969.

PROFESSIONAL AFFILIATION:

Registered Quality Engineer, License No. QU805, State of California.

Member of Subcommittee 8 of the Nuclear Power Enginee ring Committee of the IEEE Power Engineering Society responsible for the prepara-tion and revision of the following 3 national Q.A. Standards:

a. IEEE 498 (ANSI N 4 5. 2.16) : Requirements for the Calibration and Control of Measuring and Test Equipment used in the Construction and Maintenance of Nuclear Power Generating Stations.

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PROFESSIONAL AFFILI ATION: ( Con td)

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b. IEEE 336 (ANSI N45.2.4): Installation, Inspection, and Testing Requirements for Class lE Instrumentation and Electric Equipment at Nuclear Powe r Generating Stations.

c. IEEE 467  : Quality Assurance Program Requirements for the Design and Manufacture of Class IE Instrumentation and lectric Equipment for Nuclear Powc- Generating S ta tions .

I am currently a member of the IEEE Ad Hoc Committee which recommended the issues to be addressed in the development of a standard relating to the selection and utilization of replace-ment parts for Class IE equipment during the construction and operation phase. I am also a member of the work group which will prepare this proposed standard.

PERSONAL DATA:

Birth Date: 7/08/37 Married; three children Health: Excellent PUBLICATIONS AND TESTIMONY:

1. In-Core System Provides Continuous Flux Map of Reactor Corer, -

R .B . Hubbard and C.E. Foreman, Power, November, 1967.

2. Quality Assuraace: Providing It, Proving It, R.B . Hubbard, Power, May, 1972.

3. Testimony of R.B. Hubbard, D.G. Bridenbaugh, and G.C. Minor before the United S tates Congress, Joint Committee on Atomic Energy, February 18, 1976, Washington, DC. (Published by the Union o f Concerned S cientis ts , Cambridge, Massachusetts.)

Excerpts from testimony published in Quote Without Comment, Chemtech, May, 1976.

4. Testimony of R.B. Hubbard, D.G. Bridenbaugh, and G.C. Minor to the California S tate Assembly Committee on Resources, Land Use, and Energy, Sacramento, California, March 8, 1976.

l 5. Testimony of R. B. Hubbard and G.C. Minor before California S tate Senate Committee on Public Utilities, Transit, and Energy, Sacramento, California, March 23, 1976. *

6. Testimony or R.B . Hubbard and G.C. Minor, Judicial Hearings Regarding Grafenrheinfeld Nuclear Plant, March 16 & 17, 1977, Wurzburg, Germany.

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PUBLICATIONS AND TESTIMONY: (Con td)

7. Testimony of R .B . Hubbard to United States House of Representatives, Subcommittee on Energy and the Environ- lll ment, June 30, 1977, Washington, DC, entitled, Effectiveness of NRC Regulations - Modifications to Diablo Canyon Nuclear Units.

8. Testimony of R.B. Hubbard to the Advisory Committee on Reactor Safeguards, August 12, 1977, Washington, DC, entitled, Risk Uncertainty Due to Deficiencies in Diablo Canyon Qual'tv Assurance Program and Fa ilur e to Implement Current Nh6 Practices.

9. The Risks of Nuclear Power Reactors: A Review of the NRC Reactor Safety Study WASH-1400. Kendall, et al, edited by R.B.

Hubbard and G.C. Minor for the Union of Conc erned S cient is ts ,

August, 1977.

10. Swedish Reactor Safety Study: BarsebHek Risk Assessment, MHB Technical Associates, January 1978 (Published by Swedish Depart-ment of Industry as Document DSI 1978:1).

11. Testimony of R .B . Hubbard before the Energy Facility Siting Council, March 31, 1978, in the matter of Pebble Springs Nuclear Power Plant, Risk Assessment: Pebble Springs Nuclear Plant, Portland, Oregon.

12. Presentation by R.B. Hubbard before the Federal Ministry for Research and Technology (BMFT), August 31 and September 1, 1978, Meeting on Reactor Safety Research, Risk Analysis, Bonn, Germany.

13. Testimony by R.B. Hubbard, D.G. Bridenbaugh, and G.C. Minor before the Atomic Saf ety and Licens ing Board, September 25, 1978, in the matter of the Black Fox Nuclear Power Station Construction Permit hearings, Tulsa, Oklahoma.

14 Testimony of R .B . Hubbard before the Atomic Safety and Licensing Board, November 17, 1978, in the matter of Diablo Canyon Nuclear Power Plant Operating License Hearings, Operating Basis Earth-quake and Seismic Reanalysis of Structures, Systems, and Com-ponents, Avila Beach, California.

15. Testimony of R.B. Hubbard and D.G. Bridenbaugh before the Louisiana Public Service Commission, November 19, 1978, Nuclear Plant and Power Generation Cos ts , Baton Rouge, L ou is iana .

16. Testimony of R .B . Hubbard before the California Legislature, Subcommittee on Energy, Los Angeles, April 12, 1979.

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} AND TESTIMONY: '( Con t d)

! ( ) PUBLICATIONS

, 17. Testimony of R.B.-Hubbard and G.C. Mino r before the Federal Trade Commission, on behalf of the Union of Concerned S cien tis ts , Standards and Certification Proposed Rule 16

CFR Part 457, May 18, 1979,

18. ALO-62, Improving the Safety o f LWR Power Plants , MHB Technical-Associates, prepared for U.S. Department of Energy, Sandia National Laboratories, September, 1979, available from NTIS.

i 19. Testimony by R.B . Hubbard before the Arizona State Legislature, Special Interim House Committee on Atomic Energy, Overview of Nuclear S af ety, Phoenix, AZ, September 20, 1979.

20. "The Role of the Technical Consultant," Practising Law Insti-tu t er program on " Nuclear Litigation," New York City'and Chicago, November, 1979. Available from PLI, New York City.

l 21. Uncertainty in Nuclear Risk Assessment Methodology, MHB Technical j Associates, January, 1980, prepared for and available from the i

Swedish Nuclear Power Inspectorate, Stockholm, Sweden.

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22. Italian Reactor Safety Study: Caorso Risk Assessment, MHB Technical Associates, March, 1980, prepared for and available from Friends of the Earth, Rome, Italy.

23. Development of S tudy Plans: Safety Assessment of Monticello and , Prairie Island Nuclear S tations , MHB Technical Associates, 3

August, 1980, prepared for and available from the Minnesota-

) Pollution Control Agency .

24. -Affidavit of Richard B. Hubbard and Gregory C.- Minor before j.'

the Illinois Commerce Commission, In the Matter of an Inves ti- ,

gation of the Plant Construction Program of the Commonwealth Edison Company, prepared for the League of Woman Voters of i

Rockford, Illinois, November 12, 1980, ICC Case No. 78-0646,

25. Systems Interaction and Single Failure Criterion, MHB Tech-nical Associates, January 1981 prepared for and 11 from the Swedish Nuclear - f ower inspe cto rate, S tock$o$m,able Sweden.

26. Summary of Emergency Response Planning Criteria for-Regional

] and Local Authorities Near Nuclear Electric Generating Stations, 3 MHB Technical Associates, June, 19 81, p rep are d for and available from Friends of the Earth, Rome, Italy.

, 27. Economic Assessment: Own e r s h ip In te res t In Palo Verde l Nuclear Station, September 11, 1981, p rep a re d for and ,

available from the City of Riverside, California.

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PUBLICATIONS AND TESTIMONY: (Contd)

28. Systems Interaction And Single Failure Criterion: Phase II O Report, MHB Technical Associates, December, 1981, prepared for and available from the Swedish Nuclear Power Inspectorate, Stockholm, Sweden.

29. Testimony of Richard Hubbard and Gregory Minor on Emergency Response Planning, Diablo Canyon operating license hearings before ASLB, January 11, 1982.

30. Statement of Richard Hubbard before the U.S. House Sub-committee on Energy and Environment concerning QA program breakdowns, November 19, 1981.

31. Testimony of Richard Hubbard on Quality Assurance, South Texas operating license hearing before ASLB, prefiled June, 1981.

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ATTACHMEfiT B i

i RATIO OF HOSGRI/DDE ACCELERATIONS AUXILIARY BUILDING

(Presentation of Dr. Henry Kuo of NRC Staff .

to the ACRS Subcommittee on Diablo Canyon, June 14, 1978) 1

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A O O x.,

o Auxiliary B;jilding '

MAXIMUM ABSOLUTE HOR. 0.ii AL ACCELERATIONS TORS _IOtLAL ACCELERATIONS

~

HORIZONTAL. ACCELERATIONS DDE'-

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-- HOSGRI llOSGlil

'HOSGRI DDE - HOSGRI 'Y y,o DDE~

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/)nalysis Ratio DDE #"*I YN* 0"."(9)N.s -

(ll.-) hirsis (Q)

- (9) ~

~

l (9) NOSTij.SOUTil INPUT ^

- NOllTH SOUTij lyPUT 0.56 (44% Decigase) l 0.01 0.018 1.57 1.38 1.13 (13% Increase) 0.13 0.10 0.72 (2,8% Degrgase) 188.0 1.6S 1.96 0.84 (16% Decrease) 0.09 0.10 0.47 (53% Degrease) 163.0 1.23 1.16 1.06 ( 6% Increase) 0.06 0.11 0,.55 (j7% Decrease).

140.0 0.92 0.84 1.00 ( 9% increase) 0.05 0.0,8 0.63 (37% Ogcreaso) 115.0 0.74 0.62 1.19 (19% :ncrease)

'100.0 ,

EAST. WEST liiPUT EAST WEST INPUT 0.01 1.45 1.10 1.31 (31% increase) 0.12 For DDE accidental 188.0 1.66 2.40 0.69 (31% Decrease) 0.07 cccentricity of 5%

5 163.0 .

1.15 1.60 0.71 (29% Decrease) 0.05 rea s'used.

- g -140.0 0.77 (23% Decrease) 0.84 1.08 0.03

\ 115.0 0.74 0.93 ( 7% Decrease) 100.0 0.69 Auxiliary Building C

MAXIMUM ABSOLUTE VERTICAL ACCELERATIONS HOSGRI DDE* I' '

Analysis Ratio DD U Elev(ation Analysts

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(fl.) ~ (D) 0.27 2.96 188.0 0.80 .

f 0.27 2.44 .

163.0 0.66

/ 140.0 0.59 0.27 2.18

/ 115.0 0.56 0.27 2.07 - . . &

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/ 100.0 'O.54 0.27 2.00 '

/ **140.0 1.45 0.27 5.37 -

. n sed in design - $

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• No vertical dynarnic analysis was rnade for the DDE; % the peak ground accolorat on was u ]

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UNVERIFIED PG6E SKETCH WHICH

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{ LED TO MIRROR DIAGE DESIGN ERROR i

Note: Signature of the PG6E engineer who signed the transmittal was removed by PG6E.

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l ATTACHMENT D 1970 QA PROGRAM AUTHORIZATION FOR DIABLO CANYON UNIT 2 O

ATTACHMENT D f* .

b) OUALITY ASSURANCE PROGRAM AUTHORIZATION b

The Pacific Gas and Electric Company's policy is to provide safe and reliable power by designing, constructing and operating its f acilities with a high degree of quality. Adherence to this policy assures that the PG6E's activities do not cause an undue risk to the public.

In the case of nuclear power plants, we believe that to attain a high degree of quality it is essential that a quality assurance program be

, prepared and followed. Accordingly, the PG6E Nuclear Unit 2, Diablo Canyon Site, will be designed and constructed in full compliance with the Quality Assurance Program developed from the plan described in Appendix G of the Preliminary Safety Analysis Report for Unit 2. The policy for the Quality Assurance Program is set forth in i lume I, PG6E Quality Assurance Manual.

The PG6E management is fully committed to this program and I

($ hereby direct that it be implemented by those persons responsible for any aspect of the design and construction of the plant.

X JOHN F. BONNER Jartsry 1970

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O ATTACHMENT E PGSE CONTRACT AMENDMENT OF JULY 12, 1978 TO JOHN BLUME AND ASSOCIATES O

ATTACHMENT E E!;GII;EfRI!;G SERVICES File I;o. 420 GM 5234 VP&GM 69942-I  ;,;, j : ,, , ,

Mhn A. Blumeg HMSMdM *

+ n p*

URS/ John A. Blu:ne & Associates RECEIVED 130 Jessie Street CENTRAL FILES San Francisco, CA 94105

Subject:

Hosgri Seismic Evaluation

Dear Dr. Blume:

ENGINEERING DEPARTMENT This letter will confirm that you will continue to be retained as a consultant by PG&E for Hosgri Seismic evaluation at our Units 1 and 2 -

Diablo Canyon Site.

The maximum costs shall be increased from $2,800,000 to $3,500,000 and shall not be exceeded without my specific authorization.

All further work under this contract shall be conducted in accord-ance with the URS/Blume Quality Assurance Manual dated December 16, 1977.

During the life of this contract, no changes shall be made to the Manual without prior approval by FG&E.

This agreement is subject to the reporting requirements of defects and noncompliance under the provisions of Part 21 of Title 10 of the Code of Federal Regulations (10CFR21).

Please sulanit your invoices to:

Pacific Gas & Electric Company Attention: Mr. E. D. Cogswell 77 Beale Street, Room 1917 San Francisco, CA 94106 f Please refer to the project name and VPt4M 69942-I on your invoices.

Sincerely, dI ,,

(

Ag' ./ (g*- 'j? / / tsice.m r. v. ruun RVB: EPW: VJG: MPH:d j bec RVBettinger J0Schuyler EPWollak GClenfestey CVRichard s t RPWischow E-1 AT1omae

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EDCogswell

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1 ATTACHMENT F

, INITIAL CONTRACT BETWEEN 1r JOHN BLUME 6 ASSOCIATES AND PG*3 i

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RELATED TO DIABLO CANYON DATED 10/24/66 i

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ATTACHMENT F PACIFIC GAS AND ELEC* !C COMPANY eu. . .

COPY c . a scs,v s o ,008 -

DBN AAg j (7 WJP lS OCT 2 G 1966 SIN l

1CW i.E B WRF DEP A RTU*NT OF A t.t C 1.E W C N C l *J C c N O MAL FILES SCitVICES LOS8Y Octcber 24, 1566 Jdr2 A. Bitre & Accociates, EnGinocrs 610 P.xnni Street Can Preneiceo, califrnia 54105 Attentim !b. Rolcral L. 9A".x C; met:tive Vice Prcsidcut Gentlern:

Y:n nro cuth :ri.a1 to procccd on the #irst I2Vico cf your in-ver:tiCctima c:: carthqttcc crycure tand decil") criterio 1:1 c xnection with cur pr7pxcd Ebla C pm !!ttelccr P.ver PLmt, no vet f.ath in your pr.pxal Gitc.xi Octcber DJ, lia.'d.

J Irvdcen chsl:1 be directed ta 15. B. it. Chael:cifcrd, Chief Civil t

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Z".. cine.cr, rho will cico bc ;<ur flicial C:q=/ cmtcet f x thic vrh.

15. G .::dra V. nicher's vill clac c: cict ll m. N urderctar.d that y22 Icvc icrfr:rd croc vart clrecQr c;x1 thic rz:y be tined to un purcuant to this letter.

Very truly ycuro,

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• g E U:bjd bee: JF3atmer E"..hn cicif. rd i Dr.Rlly Lay,: u v.attch. /

CC..h.'1chel CVJ.ich:nt;, v.attch.

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JOHN A. BLUME

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J. P. NICOLETTI H. J. $EXTON JO HN A. BLUME & ASSOCI ATES, ENGINEERS "- '

5"^""g O. M. T ElX E' 632 HOWARD STREET

• SAN FRANCISCO, cal.lFORNIA 94105 * (415) 391 2525 October 20, 1966 Pacific Gas & Electric Company 245 Market Street San Francisco, California Attention: Mr. J. Dean florthington Centlemen:

Pursuant to your request, we are pleased to submit our proposal for furnish-ing professional consulting services for a proposed nuclear power generating plant to be located at Diablo Canyon near Avila, California. It is proposed that the work be perforned in two phases. The first phase would cover neces-sary investigations to prepare an earthquake exposure and design criteria re-port as necessary for submission by PG & E to the Atomic Energy Commission.

The second phase of work would be performed if and to the extent requested by your Company subsequent to submittal of the above report and would include evaluation of the dynamic response of structures and equipment, the integrity of which is necessary to assure a safe shutdown after an earthquake. This latter portion of the work would include but would not be limited to +'ae following:

1. Reactor building

2. Reactor vessel support 3 Containment system (if separate from buildinc)

4. Main piping 5 Ventilation stack The proposal attached hereto covers the professional services required by the first phase only, as the scope for the second phase can be better determined when more is known about the requirements for the proposed plant facilities.

He are prepared to provide the professional services as outlined in the attached proposal upon notification to proceed. It is understood that the schedule is very short and therefore work should commence at the earliest possible date.

If there are any questions on the above, please contact the writer or Mr. Blume.

Sincerely, JOHN A. BLUME & ASSOCIATES, E !GIIEERS p_3 tu Roland L. Sharpe l llW_. g1 Executive Vice President RLS/cb .

e .

JoHH A. ELUME 4 J. P. HlCCLETil H. J. SEXTO N

"'5"^""

JOHN A. BLUME & ASSOCI ATES, ENGINEERS p%) 612 HOWARD STREET

• SAN FRANCISCO. CALIFORNIA 94105 * (415) 197 2525 D. M. TEtxEm A October 20, 1966 PROPOSAL FOR PROFESSIONAL SERVICES CONCERNIUG PROPOSED NUCLEAR POWER PLANT LOCATED AT DIABLO CANYON NEAR AVILA, CALIFORNIA l This proposal is intended to cover professional services required to make the investigations and prepare the required reports as outlined under Scope of Work. The precise scope of the work to be performed by Blume and PG & E's geological and seismological consultants will be determined by PG & E in consultation with Blume and the consultants. In order to complete the re-quired work, certain site data not usually provided in conventional geology and soils engineering reports will be required. Blume personnel vill work closely with the PG & E Engineering staff and its consultants on geological, seismological and foundation work to obtain the information required in a timely fashion. The report covering recommended design criteria prepared by Blume is usually incorporated in the Plant Safety and Design Report subnitted to the AEC. In their review of this report, the AEC of ten requires the appear-ance of the author of the earthquake report is discussions are considered necessary.

Score of Work .

As directed by PG & E, Blume shall furnish professional services for the purpose of making seismic investigations in addition to those performed by PG & E's seismological consultant as necessary to establish the seismic design criteria for the Diablo Canyon Nuclear Power Plant. Such professional services shall include all investigations in conte etion with seismic conditions as are necessary to establish earthquake design criteria. Ground accelerations for baign purposes shall be determined in conjunction with PG & E's seismolog-ic.al consultants and spectral relationships shall be established. Design criteria for earthquake-resistant design shall then be established including reconmended working-stress levels, damping, etc. A report covering the recommended design criteria shall be submitted for incorporation in PG & E's Plant Safety and Design Report to the AEC.

Schedule The professional services outlined herein shall be performed in a timely fashion as directed 1 y PG & E.

Comoensation p The professional services contemplated herein shall be cocpensated for in V accordance with the following Schedule of Charges. It is estimated that the total cost for the first phase vill not exceed seventeen thousand five hundred dollars ($17,500). The rates shown in the schedule are billing rates and include all markups. p3

• , Fac, Two .

a Schedule of Charges Classification Regular Hourly Rate Overtime Hourly Pate*

President $40.00 $40.00 Principal 25 00 33 00 Project Engineer 18.00 23 50 Senior Engineer I 16.00 21.00 Senior Engineer II 14.00 18.00 Engineer 13 00 17 00 Associate Engineer 12.00 -

15 75 Junior Engineer 11.00 14.25 Technician 10.50 13 75 Drafting Coordinator 10.00 13 00 Senior Draftsman 9 20 12.00 Draftsman 8.20 10.75 Clerical 7.00 9 25

• Overtime hourly rates shall apply for all hours worked by any employee in excess of forty hours per veek provided that such overtime is authorized in advance in writing by PG & E.

Comuuter Services IEM 1404 Computer or Equivalent Systems $100.00 per hour IEM 7094 Computer or $700.00 per hour (if PG & E's computer is Equivalent Systems used there vill be no charge for computer)

Program Rentals $150 per run Key Punching Services $ 8.50 per hour Direct Expenses Direct expenses shall include such items as long distance telephone mad telegraph, reproduction of drawings and reports or other charges di$ectly attributable to the work only, travel expenses including subsistence and incidental expenses for personnel away from the home office, and shall be charged at actual cost plus ten percent. Company or personal cars used shall be billd at ten cents per mile.

Billing Invoices shall be rendered monthly. Man hours vill be kept to the nearest half hour. Payroll records vill be available for inspection if required.

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ATTACHMENT G i

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PGSE QUALIFIED SUPPLIERS' LIST FOR NUCLEAR SAFETY-RELATED PRODUCTS AND SERVICES DATED 7/78 i

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~ "s-QUALIFIED SUPPLIERS' LIST FOR NUCLEAR SAFETY-RELATED PRODUCTS AND SERVICES bwision 8 Issued By MATERIALS DEPARTMENT aly 1978 DATE QUALIFIED (REQU ALIFIED)

SUPPLIER FACTORY ADDRESS PRODUCT / SERVICES Q. A. SPEC. DATE OF EXPIRATION Anderson-Greenwood 5425 South Rice Avenue Valves and Valve Parts, 36" Max. SP-A-2 1/09/76 Houston, TX 77036 Dia. 8/04/78 "

Safety Valves, 36" Max. Dia.

Aurora Pump 840 Malcom Road Mechanical Seals and Pump Shatts 8/17/76 Burlingame, CA 8/17/78 "

Bayox, Inc. 6323 San Pablo Avenue Covered, Flux-Cored, and Bare SP-B- 1 7/06/77 Oakland, CA 94608 Electrodes; and Filler Metals for 10/27/78 Welding Bingham-Willamette 2800 NW Front Avenue Class I, 2, 3 Vessels, Valves & SP-A-2 (for N 7/18/77 Portland, OR 97210 Pump Parts, Piping Systems, Class Class i Items) 2/28/80 CS Core Supports. Class 2 & 3 SP- A-3 (f or N o Tanks Class 2 & 3 Items) b Borg Warner, Nuclear 7500 Tyrone Avenue ASME N-Class I,2,3 Valves, Valve SP-A-2 (for N 4/01/76 Valve Division Van Nuys, CA 91t #09 Parts, Vessels, Vessel Parts, and Class I & 2 Items) 10/27/78 Piping Subassemblies: 24 inch max. SP-A-3 (for N dia.,6 inch max. thick Class 3 lterns)

Bostrom-Bergen 4700 Coliseum Way Fabricate Metal Products 10/05/77 Metal Products Oakland, CA 94601 10/01/79 BS&B Safety Systems 7455 E. 46th Street Rupture Disc Assembly 11/04/76 Tulsa, OK 74145 11/04/78 Circle Seal Corp. t il No. Brookhurst St. Vessels and Parts of 8" Max. I.D. SP- A-2 (for 7/09/76 Anaheim, CA 92803 and I" Max. Thickness Code Class I 8/04/78" Valves and Parts of 2500// Max. and 2)

ANSI Pressure Class SP- A-3 (for Pipe and Piping Assemblies of Code Class 3) 4" Max. Nominal Pipe Size >

N Coast Industrial Supply 1819 5 Street Steel Bolts, Studs and Nuts

• 9/16/76 Berkeley, CA 94710 9/16/78 @

N Control Components, Inc. 2567 S. Main Street N Class I,2 & 3 Valves and 5/11/78 z d

Irvine, CA 92664 Piping Assemblies 3/01/80 a

Revision 3 - July 1978 Page 2 DATE QUALIFIED (REQU ALil? LED)

SUPPLIER FACTORY ADDRESS PRODUCT / SERVICES Q. A. SPEC. DATE OF EXPIR ATION Crane Packing Co. 6400 Oakton Street Mechanical Seal Asseinblies SP-I\-0, Rev i 9/20/76 Morton Grove, IL 60053 Packing Materials and Gaskets 9/20/78**

Dale Electronics P. O. Ilox 609 Resistors 9/13/77 Colurnbus, NEB 68601 9/01/79 Dresser Industries, liighway 71 North ASME N Class I, 2, 3 Valves SP-A-2 (for N 4/07/77 Industrial Valve and Alexandria, LA 71301 Class ! Iterns) 5/20/80 Instruinent Division SP-A-3 (for N Class 2 & 3 Items)

Edwards Valves, Inc. P. O. Box 689 Valves and Parts 5/23/78 c/o Rockwell International Corte Madera, CA 94925 3/01/80 Electromax Instr. Inc. 14 loverness Drive East Electrical Relays Mfg. SP-D-0, Rev 0 7/16/76 c) Englewood, CO 80110 7/l5/73 "

Electroswitch 167 King Avenue Relays Sl>-E-0 9/13/77 Weymouth, M ASS 02188 9/13//9 Fairbanks Morse Pump Div. 3601 Kansas Avenue ASME N-Class 3 Pomps: SP-A-3 6/02/76 Kansas City, KS 66110 60 Inch Max. Diameter 10/27/73 6 Inch Max. Thickness Fisher Controls, Coraopolis, PA ASME N Class I, 2, 3 Valves SP-A-2 (for N 4/07/77, Contiaental Div. & Valve Parts Class 1 & 2 5/19/79 Valves)

SP-A-3 (for N

- Class 3 Valves)

Gould, Inc. 6300 West lioward Street ASME N-Class 1,2,3 Valves, Valve SP-A-2 (for N-Class 1/04/77 Valve and Fittings Div. Niles, IL 60643 Parts and Machined Fittings I & 2) 3/01/79 SP-A-3 (for N-Class 3) llarrision l\olt and 2731 Wilkins Avenue Steel Nuts and flolts a 7/21/76 Nut Cornpany llattimore, MD 21223 _' /21/75" G G

' Revision 8 - July 1978 O

,' Page 3 DATE QUALIFIED (REQUALIFIED)

PRODUCT / SERVICES Q. A. SPEC. DATE OF EXPlRATION StJPPLIER FACTORY ADDRESS ASME NP & Class l&2 Pipe llangers SP-A-2 3/16/76 ITT Grinnell 'Varren, Oil _ 9/08/78 "

& MC Component Supports Pipe llanger Div. ASME NP & Class 3 Pipe llangers & SP- A-3 MC Component Supports Steel Bolts, Studs, and Nuts

• 9/14/76 ITT liarper 2014 Farallon Drive 9/14/78 "

San Leandro, CA 94577

• 10/05/77 Earle M. Jorgensen 10650 South Alameda Street Special Steel Fabrication, Aluminum 10/01/79 Los Angeles, CA 90054 Forgings Mfg. Machined Fasteners SP-B-0 8/18/75 Metrix Mfg. Company Redwood City, CA 3/31/79 Pumps and Parts ASME N-Class 3 SP-A-3 1/17/78 aNash Engineering Co. 310 Wilson Avenue 8/01/79 Norwalk, CT 06856 ,

u'

• Perform Safety-Related Inspections SNT-TC-I A* 11/08/76 Nuclear Services Corp. 1700 Dell Avenue 11/08/78 Campbell, CA 95008 Mechanical Shock Arrestors NPT SP-A-2 3/15/76 Pacific Scientific Co. Anaheim, CA 8/04/78 "

Class I, 2 & 3 & MC Linear Supports N Stamp

• 7/15/76 4201 West Peterson Avenue "REGO" 13 rand Standard Commercial

! ItEGO Division of Grade Valves 7/14/78**

Golconda Corporation Chicago, IL 60646 f Wire and Cable SP-D-0 6/10//7 11ocl<bestos Company 285 Nicoll Street 2/01/79 Cerro Wire & Cable Div. New I taven, CT Temperature & Pressure Flow instr. 5/23/78 l 11osemount, Inc. 12001 W. 78th Street 3/01/80 Eden Prairie, MN $5343 4/17/78 Lamberton Road Steam Turbines and Parts l Terry Steam Turbine 3/01/80 Windsor, CT 06095

• 7/21/76 P. O. IMx 1211 Steel Nuts and Bolts Texas liolt Comp iy 7/21/78

Ilouston, TX 77001 I

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Revision 3 - July 1973 Page 4 DATE QUALIFIED (REQUALIFIED)

SUPPLIER FACTORY ADDRESS PRODUCT / SERVICES Q. A. SPEC. DATE OF EXPIR ATION Union Pump Company Battle Creek, M1 Pumps and Parts Class 1 & 2 SP-A-2 11/25/75 Pumps and Parts Class 3 SP-A-3 10/27/73 Visalia Electric Motor Shop 30517 Ivy Road Class 1 Electric Motor Repairs and 4/19/73 Goshen, CA 93227 Modifications 4/01/30 Volumetrics 1025 Arbor Vitae Street Calibration Instruments SS-B-2 3/24/76 inglewood, CA 90301 3/27/73'*

Westinghouse Electric Corp P. O. Box 355 Nuclear fuel assemblies and assoc. SP-A-0 1/17/73 Nuclear Fuel Div. Pittsburgh, PA 15230 reactor core components (design) 1/01/79 Zetec, incorporated P. O. Box 140 Safety Related Inspections Under SNT-TC-1 A* 11/03/76 Issaquah, W A 93027 Supervision of D.E.R. 1I/03/73 9

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'I'orchase Order or Contracts placed with these Suppliers require special instructions. See Supplier Qualification File for required information.

lhis Supplier's qualification has expired or is due to expire during the next quarter.

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ATTACHMENT H PGSE'S SERVICE-RELATED CONTRACTS PRIOR TO 1979 l

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---.-,---,,.----,----,--,-..--,,.,..-,,,,-,_,-,-.--,-,,.n--,-- - , - - - - - - - - - - - - . - - - - - - - - ,

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.. December D' 981

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. .:;.n .: ..l -:. ./. '; ACTS N. ..~ .-:e.N.'1..M.a.^ms

c. v :.

SEISMIC

,7.2 24} CONTRACTORBRIEF DESCRIPTION -U.7]p REFJRTf,-:N / *_

SAFETY ONTRACT lt'l.,

DATE , *: ' .

- RELATED RELATED_

, . YES NO

. OPEN/ CLOSED  ?. .. '

YES N08 YES NO

t. . X X

- . 1971 Bethlehem Steel Breakwater Permit -

v '

/. : s. . . ' .;X!:

Geological & Seismologic~a.l # . .. X W.[l

1970'.T

-_ ~mBlume',.:J. A.:. .

Investigations;..

if

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4 M. .

Sample Analysis .

X X tr .i f 1975 Crocker. Nuclear

,.- * :. . . Lab., UofC . . , .: -

sa . . -.: . ef .. .

500 kV Shunt Reactors..,,;;,:. . .t,*. : p1h. . , . . X X

.:'.:$,.;,.S!'.,.c.a;O d74.

. .1 . . .

1971

. General Electric Kydraulic Model Study . C. O

.u f)a c.- ' ~~ J .

X X Hydro Research

. ; D.'...Q, 1970

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a ..E..;  :. 6..' ;.f. n .

X X.J I,.(.' h. -%..

h[ch)4..f:::p=., . . . 41969.j.5M. Marijave &.,Jahns q..n Geologica X X Scale Models

,. . z ;. 1971  ;. n .. .

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.,,. , , .6. ,.c: . 'e..

Construction (Breakwater) . . . :of f;@. apjfModel A.:+< ::?f?B . ..

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X< . X

", ' Stoller, S. M.,,, Hydraulic Model Study'l p,,.o ,... v.e.h 'i[ '. " '

.g : 1970: , '

w . U.C. Berkeley Cadwell Reinforcing Bar X .X .' .

'-2 1970 ,( .

-Splicing Tests . n.k, 9 x /15:::.u - m ,e . . . .

.; &,.w':(.. o.r:._

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. . Y.:. ::.9. w.::.0,

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. Design-Turbine. & . B1dg'. .f..i.{td f +.M. X-.:DM' ....;; $p' .: X ,

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. Assoc. Engs. ' c. r . .'; ,  :.. .

-pr. . . .. y Study Damage to East i U X X

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Omar J.~ 1.illevang , .-

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.js Breakwater

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Turbine Building-

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1980j. URS/ John A. Blume

VB'[sLyr. ' ,3-23-79

' T' '

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& Assoc. Modifications -

FC!s Ltr.. ,1977 WRS/ John A. .Blume .. Seismic.Research.Pmgram X RO x

-r-IN . M.: .,i[e10-E7-77(5-6-80)..& Assoc. . . .

_. X X

.: c . c

FC's*Ltr, 1975 .9 Hydro Research Hydraulic Model Study

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J.f.2-5-69'

'. H Science >

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PREUMINARY b

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s' n F..".;. ,,\;:.:m;[C ..gt. SERVICE RELATED  :

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sREPORT '

...' SAFETY SEISMIC DATE 4..~ CONTRACTOR

.ONTRACT N0. . "",'NN. :y!4YESNO RELATED RELATE [i me : ;. . / OPEN/ CLOSED' .' ~

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YES NO YES NO

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i 2-27-75 1976 z David 'W. Rice, Jr. Larval Fis.h Taxon'omicM.+. .. ..

i.:$:)s th. tr.'.U X $, . . . . . . X F,:M's .. Ltr. .:

. .; , .jP . 9. ' .fn.e '; '

.' s. Identification

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• X TH's Ltr. . 4-25-75.:1981 Winton H. Frey Final Planting Plans ,

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'(Mason L.. Hill . Geological.& Seismolog'icaTW 'X X:

TH's L.tr. ~ 4 75 ~ 1978 -

>. .. < ~ Report i . sY.

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.. X J Soi.1-Structure dntera'ctiohs.) .W.?." / -

X T.,.H 's T. , i. r .T.). 12-15-76'

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.Hosgri . Sei smic Evaluation.'i:4.,.. ....4 ,a,.;.b,.

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?!1's.'Ltrg,'3-10787..

.2 1979~ ~ Mrofessor' Ray W. i Seismic. ResponsETnalfsis');h.%y. ..

h.,. .',: M 2- . J Clough, Sc.D. of Storage Tanks . i

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F!!'s L'tr....

12-24-76.1981 , fD r..C.. Allen Cornel) Seismic Risk Analysiss . " .

Seismological Consultant ,. X i . X'

. iDr.: Bruce. Bolt. ..

FIFs'.~Ltr.

. 7-19-77 t ' Geological Exploration ~ ..? v X 1X

. ' Richard H./Jahns 3HS's.:Ltr.5

... n; . 9-23-66 1972 - .

~

X X Assessment of waves on 6. .. . ~ , . . ,.~

lWS's';Ltr.V-.6-18-68r1974 Omar J. Lillevang .

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intake & discharge. . - - ,

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structures & feasibility: 1.

.- ... .t

.+

of breakwater development. '

]? l, , Y.1 X 3WS's' Ltr. ;.' 6-12-76 .1976 . Sylvia Murray . ' Zoo Plankton & Phyto X '

Plankton Taxonomy

[.f': ' h.j - 2 .

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- ~

W< O . C .:.'. . . _2 . j.,. ,. t .

. PRELW gRY

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.... .- Dece A v , 1981 t

SERVICE RELATED CONTRACTS ,

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H ,

,:; 24 ;

),- Q :,_4 ,j;?

CONTRACTOR BRIEF DESCRIPTION ~ A' ' REPORT . SAFETY' SEISMIC CONTRACT NO. DATE

~'.' YES NO ~- RELATED RELATED

~

." OPEtuCLOSED -

~

', YES tio YES fi0 j ,

,JOS's Ltr., 3-17-68' 1977 - Omar J. Lillevang Design of Breakwatersl ' . .

X X X

IJ05's Ltr'. - 3 68 KMC, Inc. Action Plan Utility Group. . , .. .

X

~ V -

q. .

.' 1980 ,

Utility Group Report of

~

JOS's Ltr. 2-28-80 1981 -NUS Corporation

- - Investigation -

y . .. - n X X Westinghouse Electric Westinghouse Owner's Group 12-29-80  :

JOS'sLti.(*

'. I:

- < 1980

' Corp. -

X X DOS's',Ltr,' , 5-6-81 .

. 'KMC, Inc. Qualification of Reactor -

i

'[

.,.; ,1981 i

X X

q 'JdS's Ltr.' v.512-5-80 -

' Porter Consultants. Emergency Planning -

1980 Inc. Consulting Services ,

Administer Typical NRC-Type X X lJBH's Ltr'.:.. 12-4-80 1981 .-Qualification

$h.' E h.I kd', . . h . d ,;~- . ..d .. Evaluation System Reactor Operator .

-[~ . j ,,

- 9 g --

.f . s ..

' 2 Corp.'(QES) Written Examination' Consulting Services Verbal Agr. 1974 Nuclear Environment Services (% 1 .

X X 22-A-0741-0 1972 Bechtel Constr. of Scale Model-Containment Vessel ~

Eng.-7142 1972 U.C. Berkeley . Plate Bucking Testing in X X Reactor Cavity & Spent Fuel' ,

'.'s: t' ^

e '. 4 Pool Stainless Steel -

?!N.

r..

...: ?

Liners at DC 1 ->

t

' g. .. .

z..;

, i. ; ..

_3

. > ~ . - .. .

PREl.lMINARY

\ ,. . .

. . ~

[ December 5,1981 -

M* -

SERVICE RELATED CONTRACTS-

y-i; .

~ :

.y *l- '

C'ONTRACT NO. DATE CONTRACT 0P. BRIEF DESCRIPTIOrdIb b ,'hEPORT SAFETY SEIS!41C OPEN/ CLOSED . YES NO RELATED RELATED

.. . ~.y  ; YES N0 YES ti0

'i,[$g.-7138, 1972 U.C. Berkeley , Testing o'f Contain ent'IdiE. ' E.

~

fi[' # Structure Reinforced Concrete Beam i JFB's Ltr.-65 1968 Wiegel ,- R.L. Ocean Engineering'-Cooling X X Water System

~

JJ5 WilR, 7-1-66 1969 H. E. Cramer Meteorological Consultive

. ~ Services JFB's tr.-66 1970 .:sSmith &,Bentoff. . Seismological Consultants,' ,

X X

'Eng. 13-66 1969' Edward & Fred Harmon Marine Transportation X X

'Eng. 17-66' 1972 Marine Advisors Oceanographic Consulting ' X X

Y Service u-

Eng. 2 1967 Bechtel - Nuclear Fuel Cost Study . X X 3Eng. 3-67 ,('l

. ~t 1970 - ' Wurster, Bernardi Architectural Consulting -

X ,

X

. . & Emons, Inc. Services

-lEng. 4 1972 Pmntiss French Landscape Consultant X X

'Eng. 7-57 ,-1967 - White, James T. Marine Transportation X X Services Eng. 10-67 1967 Spicker, Robert Marine Transportation X X Services Meteorological Instrument. X I X

' 5ng. 11-67 1970 Montedero' ,

2- Corporation Maintenance .

Towill Corporati.on Ocean Topography X X Eng. 15-67 1969 2-14-67 . , - 1969 Dr. Edinger Temperature Studies X X Ltr.-67 1970 Frey, W. H. Soil Collecting ,

g 2 - - .

PRE GARY

.. - DecembeO1981 g,,.  :. - SERVICE RELA _TED CONTRACTS ,, l. ,

, *^ . .

..t. '}.;ff y :-

t- * . .

DATE CONTRACTOR BRIEF DESCRIPTION . ';' . REPORT SAFETY SEISMIC CONTRACT NO. ~~~ RELATED YES NO RELAIED_

OPEN/ CLOSED YES N0

. YES N0 Salo, E. O. Radioecological Consul ing -

hDW's Ltr.8-13-67 1974 Services '

X X JDW's' Ltr. 9-68 1971 'Berrigan, Paul D.. ArchitecturafServices.'lH/'.

~ '

Exploration of ~ X X 4

Eng. 10-68 1972 ,

Central Calif.

i Archaeological Archaeological' Sites Eng. 11-68 1968 . . . -(Harding-Lawson  ;. Soil. Investigation - X u X.

Associates

.- 3 ,

Eng. 12-68 .1969 Central Coast Lab. Exploratory Drilling .

X X Reception Center at Visitor's ,

Eng. 26-68 ' Closed S.M. Stoller Fuel Contract Reviews .- ': 'J. X X~

'

• Associates .

Westinghouse Engineering Consulting X 'X 5-5-69 - . 1973

• 1 '

Services Prepare QA. .

s; " '

Plans ,

5-6-69 1971 PC Exploration Rock Drilling, Placing of Explosive Charges &

Blasting Near DC 1 5-11-69 Kennedy, U. H. Periodic Audits & Inspection X of W (Tampa. FL) Pit. QA Procedures for DC 1.& 2. , ,

..?..?. .

" Steam Generators & DC 2 j){. ,

P - Pressurizer Geological Exploration X .X 8-11-69 Dr. Jahns. ,

. PREUMINARY

. December 5, 1981

SERVICE RELATED C0t[ TRACTS _ . ,

ly . ,

BRIEF DESCRIPTION REPORT SAFETY SEISMI'C CONTRACT NO. -DATE CONTRACTOR ~ ~ ~

YES NO RELAT,ED RELATED OPEN/ CLOSED YES NO

~ YES NQ~

X X 10-28-69 1972 Dept. of Fish & Game Ecological Study . -

X X 5-31-69 1971 Bellopsahi, Pietro Design Type AVL & HVT Towers Oceanographic Studies X X 5-4-70 1974 Calif. State Polytechnic '

dicodward-Clyde ~ Seismic Surveys. X X.

5-6-70. 1974 10wney & Kaldveer . Foundation,: Soil &.; X,

.5~9-70' 1971 .

,. Geological Investigations;

~

5-13-70 .2:.. i- .

1972 R. W. Hunt Inspection of Containment-Structure Liner, j: _. '-f .

~~ # ' '

Penetrations, Tanks &

Pipes for DC 1 5-18-70 - -

1975 Westinghouse Corp. Consulting Services .

X X

'5-19-70 , 1974 Garretson-Elmendorf Engineering Services' -

5-20-70 1974 Earl & Wright, Review Heavy Equipment Consulting Engs. Handling Activities at DCPP

. wBlume, J. A. : . Engineering Assistance; X .X.

'JDW's Ltr.-10-70

, 1972 (66158) 5-15-71 1976 Intersea Research Marine Biological Surveys X X Corp.

X-X .'

'5-25-71 -

1979 (EDS Nuclear I .. Piping Systems. Analysis 4~ ~

1975 Wildlife Associates Terrestrial Ecology Studies X- X

[i0-4-71 5 _

X .X JDW's Ltr,-71 1973 .Blume, J., A. . (67000) . Seismic Testing of s Electrical Equipment c.c.

. s

~

g - -

_ PREL' GARY

.' 7 ... .

' December fjl981 '

1

- SERVICE RELATED CONTRACTS - , ,

'.. :' .,.n...

-U.; . . .. ,  ;

_. .~ ,' .' * . ;. SAFETY SEISMIC BRIEF DESCRIPTION.' . . REPORT

0NTRACT NO. DATE CONTRACTOR

~ J l. :H ~ES Y NO RELA]TO RELATED OPEN/ CLOSED , .

YES N0 4ES ho

. . .~ -  :~

X X 3-24-72 1972 Roberta S. Greenwood Publication'of ' M '? .

Archaeological Report .

A . ,.; X X 5-1-72 1974 Garretson-Elmendorf Engineering Services . -

X X 5-9-72 1975 Jim J. Stillman . Marine Transportation .

Services X

5-13 -72 1974 f Robert V. Spicker Marine Transportation X Services

'.!/ ' ', ,

X 5-16-72 '1974 Smith-Emery Co. Inspection Services -

, Geologic Investigationst ' .X X.

g Earth Sciences 5-25-72 1972 Associatest

{'22-243-72

'. IQ -

Jersey Nuclear Co., Nuclear Fuel Management ,

X X I' -

. . Inc.

Environmental

& Computing Assistance. ,

Cooling Water Studies X X 1977.'

'. ' ..n 5-l-73.

- t.*;; .

Quality Analysis

• ~~

W' ,

X LFE Corporation Trace Element Analysis on X.

5h-73 1977 ,

!- Sea Water Samples.

X X 5-8-73 1979

'Garretson- Engineering Services Elmendorf-Zinov-

- Reiben, Architects

~

- & Engs. ,

. Calif. Poly. State Marine Biological Studies 'X* X hi16-73 .- 1973 Univ. Foundation l . . .

- PREUM3 NARY

December 5, 1981

?, S_ErfICF (IELATED _C_0,NTRACTS. , , ,

DATE CONTRACTOR BRIEF l..'"RIPTION' 'd's. REP' ORT SAFETY SEISMIC CONTR CT NO. ' ~~"-

• YES NO RELAIED REL%TED OPEN/ CLOSED '

.. YES N0 TES N0 5-17-73 1979 Nuclear Services Engineering'.Inalysis 'of I- ~$ X [j)'X /

Corp. Postulated Pipe Failures

- at DC 1 & 2 5-26-73 1974 Amco Steel Corp. Thermal Analysis of ,

'X X Machinery . Containment Penetrations

,. Equip. .

5-28-73 1973 State Dept. of Fish . Marine Biological Studies X X

& Game Patrick Whittle, Prepare Environmental X X 5-55-73 1979 -

' ~

Technical Specifications

. . Associates 5-56-73 1979 Stafco Associates Licensing Consulting X X Services s-60 -73 , 1977 T Bolt, Beranek & 0ffshore Geophysical'.Surve'y! .

X .. ~ X

Newman, Inc.

5-2-74 1977 Robert V. Spiker Marine Transportation X X 5-5-74 1974 ,' TEDS Nuclear. ~ Design Review Services. X X 35-8-74 1975 Ralston & Dwyer, Design Consulting Services Consulting Eng. in Connection With Electrical & Security

~

Systems at DC 1 l l5-12-74 -/ 1974 The Regents of Thermal Physical Medeling ,

X X i of UCBerkeley. Studies ,

.X I.5-30-74 1975 tAquatronics.. Marine Geophysical Studies. X j

,. 'dnternational, Inc.; .

.$. . 8- PRL(NARY

- Decc.W ,1981

,c .  !

- SERVICE RELATED CONTRACTS. ,

BRIEF DESCRkPTION I REPORT . SAFETY SEISMIC i CONTRACT NO. DATE CONTRACTOR ~ ~ ~ ~ ~

. , YES N0 _RELATED RELATED l OPEtyCLOSED

- YES N0 YES NO 1 .

Infrared & Traced Dye - X X 5-40-74 1975 Battelle-Northwest 3;' Studies .. 73 Seismological Error. - -- A, 5-42-74 1977 .Teknekron, Inc.

Analysis of Local

- Earthquake Hypocenter- -

Determinat' ions at DC 1 &'2 ,

X 5-43-74 1975 Kaiser Engineers Engineering Services X X

5-51-74 1976 Tetratech, Inc. Wave Response Calculations. . X Remote Sensed Dye Test Data X X 5-52-74 1975 Battelle-Northwest ,

Nuclear Services The Effects of Postulated- .~ X , h A_.>M 5-5-75 1981 l 1 Corp. Pipe Failures ,

Teknekron, Inc. Earthquake Source Modeling X X(

5-8-75 ~1977 5-9-75 . EWestern Geophysical- Offshore Seismic.Surveygf X ..X-

. 1975 ,

Co. of America ~ Data Marine Biological Studies X X 5-11-75 1979 State Dept. of Fish

& Game Discharge Structure Model X X

5-13 -75 1981 ~ Hydro Research Science Studies X X 5-19-75 1977 Tera Corporation Environmental Stu' dies 5-22-75 1978 Nuclear Services Feasibility Study of -

X X

Corp. Development of a Reactor Vessel Inservice Inspection Tool .

W

- PRELIMINARY

i

~

. December 5,1981 SERVICE RELATED CONTRACTS * , . . . - f.. .

![.

• ~ - .

.?..,.. .

BRIEF DESCRIPTION .' ' REPORT SAFETY SEISMIC CONTRACT NO. DATE CONTRACTOR ~

YES NO. RELATED RELATED_~

OPEN/ CLOSED YES NO 4 .

YES NO-

\* -

Ecological Studies . 7, . X 5-2-76 1976 Intersea Research .

e: X Corp.

X 5-5-76 1976 Ancrican Aerial Intra-Red Aerial ,

X Surveys, Inc. Photography Program Management. X X 5-7-76 1976 Kaiser Engineers

j. Division of Engineering & Consulting

..- . Kaiser Industries Services D' Corp.

1 X

5-12-76 -

1981 -1.ambert & Company Design & Fabricate a Reactor X

' Vessel in Service f-22-76 1977 The Regents of the X-Ray Diffraction Analysis .

Univ. of Berkeley In Connection With Compound.

g~

Identifications of DC Core ,

'[

Sediments Corrosion Product Sample X X DVK's Ltr.-76 Dr. Normal M.

Hodgkin Analysis Micrographics 5-16-77 1977 4 Earthquake iSeismic'Re-Analysis of X X

..' Engineering System, Piping Systems & Design.

Inc -

.S-26-77 1977 State of Calif. Dept. Marine Biological Studies . X X of Fish & Game

'P .

.!.x.[-

NARY g ,

1o_

PRE'

O Decerr-O-s,1981-

- SERVICE RELATED CONTRACTS

.,.r.

• SAFETY SEIS$1C CONTRACTOR BRIEF DESCRIPTION' '  : REPORT CONTRACT NO. DATE RELATED YES NO RELAT.E0_

. OPEN/ CLOSED - YES NO~ ~ES Y N0

Long Term Seismic Reanalysis X .. X 5-48-77 1977 Westinghouse Electric Co. of Containment a Reactor.

Coolant System, Interface .

With Blume Assoc..

X 5-49-77' 1978 Science Applications, Earthquake Initiated

- Inc. Associated Evaluation at a.- . ,

~~, DC 1 & 2 X X 1977 Sygnetron Security Consulting Services 5-54-77 SourceModeling5tudies X X 5-58-77. 1977 .iTera Corporation X X 1981 Garretson-Elmendorf- Design Security Building

_5-59-77

,2- Zinov-Reibin, Architects & Engs.

X X NUS Corporation Aircraft & Hissile Impet.- .

[5-60-77. 1979 -

Studies X X 5-61-77 1977 iWyle Laboratories -Seismic Re-evaluation ofc Safety Related Electricali .

Equipment . ~

Design Review of Piping. 'X X 5-65-77 1977 (EDS Nuclear .

Anchors Seismic Studies X .' X 5-66-77 , 1977 !Wyle Laboratories '

X

'1979 gApplied. Nucleonics 'Scismic Testing Studies X .

1668-77 S; .

P Co.,Inc. .

X X 1979 Battelle Mesorial Aerial Photographs 5-81-77 -

Institute . ,

PRELIMINARY

'- 11-

i , ,

December 5, 1981 SERVICE RE_ LATED CONTRACTS ..

~

% , c :. .

ONTRACT NO.- DATE CONTRACTOR BRIEF DESCRIPTION- ' - Y REPORT. SAFETY SEISMIC *

~~~

~ '

OPEN/ CLOSED .'? YES tiO_, RELATEp, RELATED

. . YES N0 YES f40 77 1980 . Applied Nucleonics / Seismic:Testin@'";f . .',' - ,' X X Co.

77 1980 Westinghouse Electric Analyze Reactor' Vessel X. ,X Corp. Support System 78' 1978

~

NUS Corporation Review & Modify Fire X

.h X u

Protection Plans r-217-78 -

1981 'KMC, Inc. ' Assist with Fire Protection X .

X Testimony -

i.EDS Huclear, Inc. Furnish Valve Qualification X 1980 X i-21 -78

? Program ,

i$i-78 1980 Jieneral Electric Co. Seismic Design X X Hodifications .

-28 . 1980 .Oly]e Laboratories

Perform Valve Operability ,

X X

. ." Tests i-31 -78 , 1979 NUS Corporation Decommissioning Study, X X Phase 1 i-35-78 1978 -

NUS Corporation Spent Fuel Storage Capacity X X Fire Protection System ,.; . . X ;X i-38-78 1981 .URS/ John A. Blume

& Assoc.

~

3-39 -78 .

1980 iGrinnell Fire Design Automatic Firei. X X. X

..:, : o Protection Sys'. Sprinkler ^ stem

. l< ; .

3-40-78 1981 Stafco, Inc. Prepare "Q-i.1st" X X 52 43-78' 1978 tDr., Richard H. Jah'ns Geological Consultation X X Services * '

5-45-78 1980 General' Electric Co. Electrical Penetration X X h Consultation Services . _ _ . h

kM--ms--OsMgaJa,L-mni.me-> sW+e'wa sJM<au44 sM M d&-+-J EAAm-- -

4+-L -+^4-- ^%ew J+3< *AA4Aa-4'- LL6 _A=AM--- - Ma, M-s A ,AmA as mm--4 6. w.~u-s,uL.--- A.ansexa b ,&

!O i

i

! ATTACHMENT'I i

4

. t l PGSE'S LIST BY VENDOR OF ELECTRICAL EQUIPMENT (

REQUIRING ENVIRONMENTAL QUALIFICATION I

1 1

l l 1 i

.i s

O

ATTACFBIENT I FILE LIST (I File Equipment No. Type IH-1 Rosemount 1152 IH-2 Barton Lot 1 IH-3 Barton 351 IH-4 Sostman RTO IH-5 W Fan Coolers IH-6 ASCO NP Series IH-7a Limitorque SMB I/C Class H IH-7b Limitorque SMB I/C Class B IH-8 NAMCO EA 180 IH-9 Valcor IH-10 ASCO Non-NP I/C IH-11 Fischer Porter 10B 0/C IH-12 Limitorque SMC-04 IH-13 Fischer Porter 50 EP IH-14 ITT Gen. Cont IH-15 Target Rock IH-16 Limitorque SMB 0/C IH-17 _ASCO NON-NP 0/C IH-18 Fischer Porter 10B I/C IH-19 ' Burns RTO IH-20 Hydrogen Recombiners IH-21 Acoustic Leak IH-22 i Barton 332 IH-23 Barton 763 IH-24 Barton 764 IH-25 Hydrogen Monitor O

s/ I-1 0712A

~

FILE LIST (Continued)

File Equipment No. Type IH-26 Victoreen IH-27 Barton 288A IH-28 Rotork EH-1 GE Penetrations EH-2 Okonite Cable, EPR/Hypalon EH-3 Raychem Cable, Flametrol EH-4 Rockbestos Cable, Firewall III EH-5 Boston Cable, Silicone /Hypalon EH-6 Okonite Cable, Tefzel EH-7 ITT Cable, Thermocouple EH-8 Continental Cable, Silicone EH-9 Boston Cable, Fan Cooler EH-10 Raychem Cable, Stilan EH-11 Raychem Splice EH-12 Conex Conouctor Seals EH-13 0.2. Gedney Conductor Seals EH-14 W Motors; RHR, Charging Motors EH-15 W Motors; SI Motors (Unit 1)*

EH-16 Okonite Cable, SKV

• Unit 2 Pump Motors removed and sent to Con Edison.

/

l l

I-2 g

l l

l 0712A t

t O

l i

l 1

I ATTACH 5!ENT J PGSE REPORT DATED 7/29/79 0F URS/BLUSIE AUDIT I

l I

l O

i

ATTACHMENT J FACIFIC GAS AND ELECTRIC C O M PANY PG .*E l n etAtt stRcti . SAN rRANCISCo. CALIFORNIA 94106 . (415) 781 4211 . TWX 910 M76LS7

• June 29, 1979 i *n 3 fir. R. F. Runge Quality Assurance tianager 1 URS/ John A. Blume & Associates, Engineers 130 Jessie Street W San Francisco, CA 94105 -

O

Dear Bob:

O

• I performed an audit at your San Francisco offices during the period of O April 2' through !!ay 24, 1979. The purpose of this audit was to confirm that URS/Blume is implementing contractual and Quality Assurance require-ments. You are aware of the problems and deficiencies.which we encountered as well as the length of time and man-hours needed to verify certain aspects of your quality program.

i While the audit report has been written, the resolution to several audit findings remains to be verified; in general, I expect them to be resolved with the implementation of Revision 7 to your QA llanual.

Some of the deficiencies in your-program may seem minor, but it is PC&E's responsibility to verify operation of your program to our satisfaction.

For this reason, I requested the assistance of several engineers in reviewing various technical aspects of your projects in order to verify the accuracy and quality of the work.

In the future, PG&E expects your Quality Assurance Program to be fully i

functional and all your project personnel familiar with the applicable quality requirements of their projects. A follow-up audit ?:ill be conducted later this year to verify the resolution of all outstanding findings.

j Sincerely, ,

rblA Y

! }i. L. Barham MLB(3692):er J-1 t

, - - , - , - , - , , . . , , - e. , ,. - -

76-jiu (14/ so)

Jr  : *a DISTRIBUTION OF AUDIT REPORTS O

In accordance with Quality Assurance Department Procedure 11.8, lf7 " Audit Reports: Form, Content, and Distribution", audit reports are to O be submitted to the Director, Quality Assurance, within 2 weeks (10 working g days) of the conclusion of the audit.

n If the above schedule is not met, explain below:

1 N .

C O Audit Report No.: ,

O o C i" 3'%t.""'" #Mr2I/97/ _

Issue Date of Audit: M N >/ 7[

Reason for Delay: bM [

Ju ud m A uzadeu. N J,

/ // y 11&fh e "hWW IL.

$'af Esaz" <J n 2&

ams n,4 tk 4& .pp - . a & .

W. Raymond J-2 g

a s ,

QUALITY ASSURA;;CE

() 11.1 x d.0 Audit of URS/Blupe 6 Associates, Engineers Sun Francisco, California Audit I:o. 91605 ND O

07 June 29, 1979

.O f

HR. B. W. SHACKELFORD:

04 Attached is the repor't of an audit of URS/Blume perfurned on C3 April 23 through !!ay 24, 1979. URS/Bluce provides consulting C3 services for seisuie qualification of structural (and nechanical) supports at Diablo Canyon Power Plant. This work includes C3 design reviews, analyses of support.systens, and new design for modifications.

CD The audit was psrforb'ed to ver,,1fy that this consul. tant was adequately inplementing its quality assurance progran on the PCandE projects for which it is responsible.

The auditors found that in 12 instances URS/Blune was not inplenent-ing contractual and Quality Assurance !!anual requirecients. Due to the significant nature of the audit findings, it was concluded that URS/Blune had not been responsive to its connitnent for a quality prograu. . Innediate action was taken to inforn Project Engineering of the findings, and to schedule corrective actions.

After the initial findings, PCandE engineers were asked to assist the audit tean to verify the quality of URS/Blune's design work.

In general, the design work itself appeared to be satisfactory.

Two computer prograns which had not been verified have sinc'e been j verified to comply with regulations.

In addition to the deficiencies identified in the URS/Blune QA Progran, deficies.cies in PCandE's interfacing activitics with con-sultants were also identified. All findings have been dccupented on l'onconformance Reports #DCO-79-EU-004, -005 and -006. Corrective actions will be verified by the Quality Assurance Departncnt.

J. A.  ::0?:D iCU/ULBarhan (3692):rr cc: DC 6 !!B - Supplier f-)

s Attachment J-3

~ - __

1 5/9/79 DISTRIBUTION OF QA AUDIT REPORTS DIABLO CANYON AND HUMBOLDT BAY - SUPPLIER N

O JRAdams 1A RCAtkins y RSBain y WRBart GSBates RPBenton O

FWBrady C

DABrand O HPBraun o JGCarroll ,

PAcrane, Jr.

JYDeYoung RDEtzler JBHoch DNKelly EBlangley, Jr.

FCMarks FFtbutz CKMaxfield PTMurray RDRamsay HWReynolds JVRocca BPSadler J0Schuyler BWShackelford (original letter)

EDWeeks JDWorthington 3-4

Audit No.: 91605 Issue Date: 6/29/79 Page l'of 2 O

PACIFIC GAS AND ELECTRIC COMPANY

. QUALITY ASSURANCE DEPARTMENT cc C

Title:

Audit of Supplier's Quality Assurance Program y) Audited Organization /

Facility: URS/ John A. Blume & Associates, Engineers 1

,, Auditors:

M. L. Barham (Lead Auditor)

C. L. Eldridge C3 Dates Performed: April 23-May 24, 1979 C3 C3 1.0 Scope ,,

URS/Blume has been providing engineering ' consulting services, including design, review, and analysis, for seismic qualification of structural supports. This audit was performed to verify implementation of all aspects of URS/Blume's quality assurance program related to their present work at Diabla Canyon Power Plant.

2.0 Conclusion On May 24, 1979, a post-audit conference was held at URS/Blume's facilities.

Twelve audit findings were presented and discussed. The following people were in attendance:

PGandE URS/Blume -

M. L. Barham Betty Anderson C. Kensler C. L. Eldridge B. Chakravartula H. Klice M. E. Lee D. Dilip D. A. Lang J. J. Benitou R. Gallagher P. K. Lum G. Hess B. Mullin K. Honda R. F. Runge K. Jay R. Yokoyama The audit findings were discussed and audit finding reports presented to URS/Blume's management. The deficiencies occurred in the following areas:

design review, internal audits, annual reports to management, organization chart, audits of Document Control Center, computer program verification, and interface procedures. Details of the audit findings are in Appendix A.

O J-5 V

. - - I

Audit No. 91605 Appendix A Page 1 of 8 APPENDIX A AUDIT ACTIVITIES C3 1.0 References

-- (1) 10CFR50, Appendix B (2) ANSI N45.2.9-1974

'S (3) ANSI N45.2.ll-1974 IO (4) ANSI N45.2.23-1978 (5) URS/ John A. Blume & Associates, Engineers Quality Assurance 7  !!anual, Revision 5, 6/9/78

?4 2.0 Docucents Reviewed .

c) (1) Training records (2) Quality Assurance !!anual revision control records C3 (3) Design calculation packages g) (4) G-Line Anchor Design Review Package (5) Specifications and design criteria --

C3 (6) Computer file registers, program logs, and certified proc, ram list (7) Project file registers (8) Document Distribution Lists (9) Design Review Reports (10) Audit Deficiency Reports 3.0 Audit Findings 3.1 Supplier Audit Findine Reports Findig Recommendation (1) Independent review of critical Obtain written direction from aspects of URS/Blume work is PGandE specifying ht'e actual required to be performed by design review completion dates a Quality Assurance Engineer for each of the projects at prior to transmittal of work URS/Blume Engineers.

to client. Such reviews have not been performed on the bulk of work prepared for PGandE.

Documents have been transmitted to PGandE, accepted for con-struction, and used to perform work at the site.

(2) Internal audits have not URS/Blume must conform to been performed in accordance its QA requirements for audit with frequency requirements scheduling; Blume must perform described in URS/Blume's an audit on each of the projects Procedure 6.1. it is working on for PGandE.

J-7 h

• ,y Audit No. 91605 Appendix A Page 2 of 8

,n Finding Recommendation (3) Requirements for annual Comply with requirements; perform program review and report program review and prepare a

__ to management have not report to management on the been met for 1978. status of the QA program including achievements as well as problems.

LO (4) Organization chart does Revise organization chart to not accurately represent accurately represent present j the present URS/Blume lines of authority and organization. responsibility.

74 (5) URS/Blume's Quality Assurance Perform regularly scheduled C3 Manager has not performed audits of Document Control Center

,, the required audits of the as required by the QA program.

Document Control Center.

es

'~

(6) Several major computer Complete documentation, as C3 programs used'for analysis of required by QA Manual, for all PGandE projects do not have computer programs. PIPESD' the required verification and (version 5.2 and 5.3) must have certification. This informa- the procedure and method used tion is required to be filed to verify.

in the user manual kept by the Computer Services Manager, and its availability is of extreme importance.

(7) Several less critical program- Complete documentation, as routines used on project required by QA Manual, for piping analysis have not all computer programs used on been certified or verified. PGandE projects. , Modify Manual The background data for to explain how some programs, MODPIP, ENVELPS, and RREPS unique to a project, are (Rapid Reanalysis) are kept by maintained and stored with the the Project Manager; however, project files and not with the the QA Manual states that the Computer Services Department.

Computer Services Manager is responsible for maintenance of all program documentation.

(8) Interface procedures have Write interface procedures to not been written or approved identify interactions between as required by Procedure 4.6. URS/Blume's office, its field forces, and PGandE. Procedures to be approved by QA Manager.

Perform audits of field activi-4 ties to verify compliance with J-8 the QA program.

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Audit N2. 91605 Appendix A Page 3 of 8 Finding Recommendation O

(9) Responsibilities, duties, and Interface procedures for field 04 authority of field personnel personnel's authority, respon-are not clearly defined in sibilities, and duties must be

~~

writing. The field's inter- clearly defined.

face with the QA Manager is e,n not clearly defined.

30 (10) Current revision numbers Revise QA Manual to require that 3, for design criteria are not the revision number (or date) of identified in writing. design criteria be entered on pq Document distribution lists the document distribution list.

have not been completed and Start filing memos which verify C3 do not include revisions to receipt of criteria by those on

,, the criteria, the distribution list.

) (11) Several sections of hand Have these calculations calculations performed on checked in the same manner as C3 the Intake Structure Dynamic the balance of the project -

Seismic Analysis have not by a qualified engineer been checked as required by who was not involved with URS/Blume's procedures. performing the original calculations.

(12) Several pages of calcula- Have the calculations reviewed tions of eccentricity for by the Project Manager and the Spectra generation are not status of calculations noted signed by the checker. on the calculation index.

3.2 Open Item Report 075-79 Finding Recommendation Safety related design Establish requirements for documents generated and adequate document storage for retained by URS/Blume are not URS/Blume generated documents.

being stored in accordance Review and take appropriate with fire protection require- action for all other con-ments of ANSI N45.2.9-1974 sultants who generate Many documents such as design safety related documents.

calculations and computer program certifications appear to be the only copy in existence.

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' . ' ' *r Audit No. 91605 Appendix A Page 4 of 8 O

v 3.3 Nonconformance Reports (1) NCR #DC0-79-EN-004 D7 The URS/Blume Quality Assurar :e Program for Work on the Diablo

_. Project has not been effectively implecented. Nine Supplier Audit Finding Reports have been issued and are being followed in through this Nonconformance Report.

50 The cause of the nonconformance was unrealistic scheduling con-I straints and lack of detail in contract documents with respect

.to quality assurance requirements.

M

. This NCR will be resolved by having technical consultants from C3 the Engineering Department assist the audit team in reviewing C3 work done by URS/Blume. The audit findings will have to be resolved to PGandE satisfaction in order to close this NCR.

O (2) NCR #DC0-79-EN-005 o -

The URS/Blume quality assurance program has not been effectively implemented and is unable to provide assurance that design work performed is adequate.

This NCR was canceled and the deficiency covered by a revision of the resolution to NCR #DCO-79-EN-004.

(3) NCR #DCO-79-EN-006 There is no clearly defined interface between URS/Blume and the Engineering Department as required by ANSI N45.2.11 ($ 2.2 & 5.1).

Additionally, the Engineering Department manual does.not presently require that design interfaces be identified and controlled between PGandE, URS/Blume, and other architect engineers as l

required by the Standards.

This NCR will be resolved by having the Engineering Department define the design interfaces between PGandE, URS/Blume, and other architect engineers. The Engineering Department manual will have to be revised to require design control measures as specified in the ANSI Standard.

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Audit No. 91605 App;ndix A Page 5 of 8 4.0 Additional Information 9

4.1 PGandE Personnel Accompanying Audit Team W R. E. Bacher M&NE

_. J. J. Benitou EQC V. M. Chauhan M&NE t.9 V. J. Ghio CE 3 S. D. Krein EQC M. E. Lee CE 1

G. R. Veerkamp M&NE N

g 4.2 Additional PGandE Personnel Contacted O D. J. Curtis o J. B. Hoch C. M. Li -.

O J. J. Kilian

~

A. G. Walther E. P. Wollak 4.3 URS/Blume Personnel Contacted B. Anderson B. Chakravartula P. Chokkalingan R. Gallagher G. Hess K. Honda K. Jay C. Kensler H. Klice D. Lang P. Lum B. Mullin C. Quan R. Runge S. Swan R. Yokoyama J-11 g

1

Audit No. 91605 Appendix A Page 6 of 8 O

b 4.4 Additional Findings by Engineering Department Several PGandE engineers were informed of the state of URS/Blume's A QA program and asked to accompany the audit team. The intent was to determine if the lack of implementation of the QA program had resulted in actual design problems and questionable results or if

'A it resulted mainly in documentation omissions.

The engineers involved in the audit were directed to critically y- examine selected portions of URS/Blume's work to verify that it meets the requirements of the design criteria.

W g Mr. M. E. Lee (C.E.) was responsible for reviewing design verification performed on the Intake Structure analysis. He verified that the o computer programs used on this project (TABS and SAP IV) were verified per the QA program. He reviewed the hand calculations and discovered C that the last three sections (of ten sections) had been prepared by the project manager, D. Lang, and not checked.

C3 -

These sections dealt primarily with answers to RC questions on Hosgri Seismic Analysis which had been prepared under difficult scheduling constraints. The project manager has stated that he will either have the calculations checked (as the bulk of the work has been) or place a statement in the file that the design review verified the calculations.

This item will be followed with SAFR #91605-10.

The PGandE Engineer checked the design review package, assumptions, dimensions of structures, computer output file, and the summary of Final Revised Design Analysis with incorporated changes from URS/Blume's Designing Engineer. He was satisfied with the work.

Mr. C. M. Li (C.E.) was responsible for reviewing design erifict. tion performed on the Turbine Building modifications. He verified that the computer programs used on this project (DRAIN 2D and SAP IV) have been verified per the QA program; in addition, he reviewed the user file and test program certification. He reviewed a random 507. of the calcu-lations for the signatures of the designer and checker. He verified that design is reviewed by at least two other engineers against the l design criteria. He checked the reviewer's calculations and feels l

that the results of the design review are acceptable. His contacts

'were H. Klice, D. Lang, and R. Yokoyama.

Mr. D. J. Curtis (M.E.) was responsible for verifying the quality l of work on the piping support review. He chose Unit I fire piping supports at random. They were support Nos. 48/35R, 56N/54R, 555/6R, l 58N/21R, and 41/57R.

I i

3 J-12 (U

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Audit Na. 91605 App;ndix A Page 7 of 8 Calculations were checked to insure that the following items were O

properly included and calculated:

(a) Design Criteria, (b) Consideration to Interfacing Disciplines (i.e., giving reaction loads),

an (c) Consideration to Gaps (i.e., space required for thermal movement and maximum space allowed to

.7 be considered a seismic restraint),

1 (d) Natural Frequency, N (e) Actual Loads and Stresses, and O (f) DE Deflections.

O It was not the intention of this review to investigate the C constructability of modifications proposed by L3S/Blume.

O It was noted that in the calculations for pipe support 56N/54R, a reducing factor for the allowable load on a concrete expansion anchor was used. PCandE pointed out that this was unnecessary and unduly conservative. URS/Blume acknowledged and stated that this practice is no longer being done. All other items were found to be satisfactory.

URS/Blume is, in general, doing a satisfactory job of documenting the adequacy of pipe supports.

Mr. V. M. Chauhan (M.E.) was responsible for verifying the quality of work performed on the piping support review. He examined the certification of computer programs and the logic banis, reviewed hand calculations, analysis used in spectra programs, and justification for assumptions used in developing the analysis.

He found two errors where incorrect data were entered into the computer for computing hanger loads. Investigation revealed that the incorrect data resulted only in a more conservative load analysis.

He reviewed the " Rapid Reanalysis of Piping Systems - Hosgri 7.5 Evaluation" which is a program developed by Blume in 1977. The user manual is incomplete and contains no certification or verification.

This engineer has several other yet unanswered questions which are documented on Supplier Audit Finding Reports Nos. 91605-11 and -12; resolution of these findings will be verified by Quality Assurance.

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Audit No. 91605 Page 2 of 2 O

t v

The deficiences identified are not minor and indicate that a problem exists in the implementation of URS/Blume's quality assurance program.

In addition to the URS/Blume deficiencies, problems were identified within the interfacing activities between PGandE and its consultants. The Quality Assurance Department has issued three Nonconformance Reports to document a these audit findings and will follow their progress to resolution.

if) 3.0 Nonconformance Reports (NCRs) Initiated During This Audit M 3.1 NCR #DCO-79-EN-004 Responsible Department: Engineering I

Description:

The URS/Blume QA program for work on Diablo N Canyon has not been effectively implemented.

Audit Findings Nos. 91605-1 through 91605-9 O have been issued to URS/Blume.

O 3.2 NCR #DC0-79-EN-005 Responsible Department: Engineering C

Description:

URS/Blume's QA program noc effectively O -

implemented and as such is unable to provide assurance that design work performed is adequate.

3.3 NCR #DCO-79-EN-006 Responsible Department: Engineeri..g

Description:

No clearly defined interface between URS/Blume and the Engineering Department as required by ANS1 N45.2.11 ($ 2.2 & 5.1). Engineering Department manual does not require that design interfaces be identified and controlled between PGandE, URS/Blume, and other architect-engineers.

Additional details of these NCRs are in Appendix A. **

Performed b :

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l Barham f u C. L. EldriGhe (f Approved by: # m ,,

W. A. Raymond Attached: Appendix A J-6 /

l

Audit No. 91605 I Appendix A Page 8 of 8 Mr. R. E. Bacher (M.E. ) was responsible for investigating the i verification and certification of the PIPESD program. He reviewed the background information for versions 5.1, 5.2, and 5.3; all revisions D*

and changes to this program have been verified through the use of benchmark problems. In three separate problems, PIPESD was compared

__ respectively against ADLPIPE, ANSYS, PIPDYN, and SAP IV; in all cases, the solutions were acceptable and the analysis documented to the

';n satisfaction of the responsible engineer.

3 It is felt that the critical review of Blume's performance on the j Diablo Canyon Project has resulted in strengthening their program and providing our Engineering Department with sufficient data to verify 04 the quality of work performed by Blume's engineers.

O O

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ATTACHMENT K l

4 EXAMPLE - INFORMAL PGSE INTERNAL DOCUMENT TRANSMITTAL i

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, ATTACHMENT K WrSNib'51 b. 0NV.l$Eh $ $.U'GhihS&;.$lGihi. Wd.. .

I 130 Jessie Street (at New Montgomery) San Francisco, C lifornia 01105 - .

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             .,,. Pacifie Gas 6 Electric Comoany Atte ntion:                   Mr.VinecTi$

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Subject:

Containment Structurc 'y -. V.4', y,;j 4 Room 2609 g Ccale Strcet, '

         '              San Francisco, CA 94106                                                                                   . Annulus Vertical Spectra I
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          ;  ,.vc are forwarding                                                                                                  Preliminary D rawi n g s ....___
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                '3          FOLLOWS :

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                        ..'l.pyJ,'s .1 F l oor R esp,ons e Spec stra */. 5M ijpygri Ba,r,thau3k e," Au;3if t 1,'1977 The r.pcetra_pyq1.c.q.                        t       ed in cur May 9 report ccusideredg'...-~                                       at             .'             -
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                 . w xu'd n1 m like to noint out that fuIIhey investigation has_.Icsulted in
                   .. in..Ercased amplification factors in the vicinity of T = 0.03 see for selected nodal points.                   This applies to both the fixed and pinned connection analyses.

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O ATTACHMENT L QA FOR URS/BLUME WORK FOR THE DIABLO CANYON PROJECT i i i O i

               -- - - - ,._--,en.---,-.- -, _ .- ---, ---- . , ___-- , ,- , __ _ _- ---- - -.--,, .-- -

ATTACHMENT L g E. URS as =w c ... ::.: m : ... t= . :s :c, :- (3 URS: JOHN A. BLUME & ASSOCIATES, ENGINEERS ii:;/R- - O 130 JESSIE STREET ( AT NEW MONTGCMERY) 1+.'3 *a SAN FR ANCISCo. CALIFORNIA 94105 C '=

                                                                                           -0 TEL: (415) 397-2525 di .6 CABLE: BLUMENGRS ;^*;{C 

if( J"-U" August 30, 1977 E.2F" FFU GECCiVED Ot.P Mr. J. O. Schuyler - NUCLEAR PROJECTS DES Manager, Nuclear Proj ects JBH DEPARTMENT p/ Pacific Gas 6 Electric Company EHM 77 Beale St- eet, Room 2645B NLZ SEP 2 1977 __ _ San Francisco, CA 94106 FILE

SUBJECT:

Quality Assurance for URS/Blume Work for the Diablo Canyon Project

REFERENCE:

URS/Blume Quality Assurance Manual

Dear Mr. Schuyler:

This letter summarizes the quality assurance procedures followed in the course of our work on the Diablo Canyon proj ect. Due to the fact that the present 10CFR50, Appendix B quality assurance requirements came into existence well after our work on Diablo Canyon was begun, it is necessary to separate our work into three parts. Since our work on the DDE analysis of Diablo Canyon (pre-Hosgri) predated formal QA requirements, we followed good engineering analysis practices including checking of all calculations, computer programs, and reports. Our work on the Hosgri evaluation of the Diablo Canyon structures employed the same thorough engineering practices and in addition included many of the formal quality assurance requirements specified in 10CFR50, Appendix B. Our present and future work on the structural modifications to the Turbine Building will employ complete quality assurance requirements in accordance with the URS/Blume Quality Assurance Manual. In general, the URS/Blume quality assurance standards require three primary areas of action during the course of a project: design control, document control, and audits. Design control requires conformance to regulations in j the following eleven categories:

1. Criteria, or Basis for Design-

a. Will be clear and complete.

b. URS/Blume Proj ect Manager is responsible for development.

l c. URS/Blume Division Director is responsible for approval and client agreement.

d. Cannot be changed informally. All changes will be made according to Quality Assurance Manual section on revisions.

Oj

2. Calculations

a. May be hand or computer, 1

l L-1

URS ~ w cuascesiseu s :s. ss=. :n :=w.:c Mr. J. O. Schuyler August 30, 1977

b. Index for all calculations will be included.

9

c. Will show all applicable criteria, assu=ptions, and refer-ences.

d. Will show enough so that any qualified engineer not familiar with the project can understand what is being done.

e. Verification of results: sign off by originator and checker.

URS/Blume Project Manager signature indicates completion.

f. Revisions according to Quality Assurance Manual.

3. Computer Programs

a. All programs used on quality assurance projects are verified programs. (In the e">.t an unverified program is used, a second program will ve run as a check.)

b. All certified programs will be logged with name, version, principal use, and personnel available for consultation.

c. In-house programs will be certified by our computer group.

d. Certification involves documentation of everything that went into making up the program (flow charts, listings, user manuals, and verification of output).

e. Outside programs: verification calculations will be available.

Where verification is done outside the company, documentation will be on file as part of the project documents,

f. Modifications: all modifications to programs will be approved by the URS/Blume Division Director. Modifications will be made only to copies of the program, never to original versions.

All modifications will be documented. Modified programs will be logged with proper version of identification after they have been certified.

4. Drawings

a. All drawings will be checked for compliance to calculations, project criteria.

b. Final approval by the URS/Blume Project Manager,

c. Index of all drawings will be maintained by the chief draftsman.

d. Revisions to drawings will be documented and follow the same checking procedures as originals.

5. Field Investigation

a. All field work will be documented,

b. Equipment and instrumentation used in the field will be cali-brated, and calibration records will be filed with the project documents.

6. Interface Control

a. URS/Blume Project Manager is responsible for establishing the procedures between URS/Blume and other organi::ations (architects or other engineers).

L-2

URS as v$s. 4. csa, pa;s g gg ;*,4. gg a, ;g; n;4 ,;.* Mr. J. O. Schuyler August 30, 1977 7 ~

  ,)

b. URS/Blume Quality Assurance Manager reviews all procedures to assure compliance with quality assurance procedures.

c. All procedures are documented.

d. Objective of interface control is to keep all involved parties informed and prevent unnecessary rework.

7. Reports

a. URS/Blume Division Director has overall responsibility for information and recommendations.

b. URS/Blume Project Fbnager responsible for preparation.

c. Revisions documented and reviewed; one copy to project files.

8. Specifications

a. URS/Blume Project Manager has major responsibility.

b. All specifications shall be checked for agreement to project drawings and other project requirements.

c. Ilandled similarly to reports.

9. Subcontractors

a. Must have quality assurance program or be subject to pro-cedures of URS/Blume Quality Assurance Manual,

b. URS/Blume Project Manager responsible for selecting subcontrac-tors,

c. URS/Blume Quality Assurance Manager responsible for subcontrac-tors' compliance to a quality assurance program. Must approve subcontractors before selection.

10. Design Reviews

a. There will be at least one review per project, depending upon project size.

b. A review will be a check of selected portions of a project to assure conformance to the project criteria and project require-ments as well as a spot check of calculations.

c. Whenever possible, the quality assurance review will be performed by qualified persons not working on the project.

d. All discrepancies will be identified and documented,

e. A design review report will summarize all discrepancies and corrective action taken.

11. Revisions

a. No revisions will be made without proper documentation.

b. Revisions and all necessary documentation apply only to approved documents.

The control of quality-related documents require that file registers or (~^ indexes be maintained at all locations where quality assurance required s- documentation is stored. The retention of all documents will be as directed L-3

URS as. i... 2 ....:,i3 w .. 31.,::,: o :. Mr. J. O. Schuyler August 30, 1977 O by PGGE. An index of all material in the Document Control Center will be maintained by URS/Blume. Internal audits will be performed within URS/Blume to verify that quality assurance standards have been properly implemented and are functioning effectively. Audit teams will be made up of members of the firm not in-volved in the project. Pre- and post-audit meetings will be held to prepare for the audit and to discuss any discrepancies found. All audits will be completely documented. A compilation of the quality assurance requirements which have been or are being met for the Turbine Building modifications and the recent liosgri evaluation work are shown in Attachment A. The measures described herein represent a summary of the quality assurance requirements specified in the URS/Blume Quality Assurance Manual. We will be happy to provide you with a more detailed description should you require. Very truly yours, URS/J0llN A. BLU)lE 6 ASSOCIATES, ENGINEERS N 'c Davi fA. Lang Project Manager DAL:bem Attachment cc: Mr. Vince Ghio/PGSE Mr. Darrell Polley/PG6E L-4 l

1 i ATTACIBENT A URS/BLUBE QUALITY ASSURANCE DIABLO CANTON NUCLEAR POWER P! ANT TURBINE BUILDING RECENT HOSGRI l DESIGN CONTROL MODIFICATIONS REEVALUATION WORK

1. CRITERIA e e

2. CALCULATIONS e e

3. COMPUTER PROGRAMS e e

4. DRAWINGS e e

5. FIELD INVESTIGATIONS e o

6. INTERFACE CONTROL e o

7. REr0RTS e e

8. SPECIFICATIONS e N/A

9. SUBCONTRACTORS e -o

10. DESIGN REVIEWS e e

11. REVISIONS e o DOCUMEE CONTROL e e AUDITS e e l

i I 1 I e REQUIRESENT MET OR BEING MET o REQUIREMENT NOT PET N/A NOT APPLICABLE l lO L-S l i

      - - - - - - , .~           -   ,-  . - - - -         - - - ,,,-           ~   n - , - , . -,- . _ - -                                            . . _ . . . . ,     --,,. _.-. -,.,..e-

1 O l ATTACHMENT M PGSE QUALITY ASSURANCE AUDIT OF DESIGN REVIEWS DATED JUNE 19, 1972 I !O

ATTACHMENT M 4 Quality Assurance Audit - p) w. n "De s i en,_It eu f eur." From !!ay 2 through May 15, 1972, an audit uas conducted by Verne E. Steen and Jens Erlingsson, both of Quality Assurance. The audit was conducted to determine if Electrical Engineering, Mechanical Engineer-inn, and Civil Engineering are conducting design revious on all Class I systcmr. and structure::, in accordance uith QA Procedure No. PP.E- 6 . For the purpose of the audit, the follouing individuals ucro contacted: Mr. W. J. Lindblad - Project Engineer Mr. R. A. Young - Electrical Endincering Mr. D. Nielson - Mr. 8. P. Wollak - Civil Engineering Mr. V. J. Chio - " " Mr.11. J. Cormly - Mcchanical Engineering Mrs. W. Ilutton - Civil Design-Drafting Electrical Engineering The audit of Electrical Engineering revealed that this engineer-ing discipline has generally not conducted design reviews on cicetrical Design Class I systems in accordance with QA Procedurc No. PRE-6. Independent checking of calculations was done on most Class I systems by an independent engineer, uho had no involvement in the design. llowever, the reviewing engineer did not preparc a report of his work to

• assure that PSAll requirements, codes and standards, functional requirc-ments, assumptions, if any, ucre translated correctly into the construc~

tion drawings and specifications, as is required in QA Procedure No. PitE-6. Electrical Engineering had, prior to the audit, developed a sys-tem design checklist for guidance in preparation of design reviews. Mr. D. Nicisen indicated that it is the intent to use the checklist as a cover sheet, such that a person technically qualified in the subject can review and understand the revieu process and be assured that portinent criteria have been satisfied without recourse to the originator. The system design checklists ucre, at the time of the audit, being filled out and inserted in the design revicu folder. Mr. Nielsen further indicated that the cover sheet and the check calculations will comprise the design revicu report and will satisfy the requirements in QA Procedure No. Pile-6. Mechanical Ennineering Eighteen mon:hs ago, Mechanical Engineering performed design reviews on the Component Cooling Water System, the Auxiliary Saltuater System, the Containment Spray System, the Diescl Generator Unit and Fuel (7 Rj M-1

l l l Oil System, and the Auxiliary Feeduater System. 1hc revicus, houever, l wcre performed on the five r.yatems when the designs were only 607. to 707 - l complete. Design revicus have not been conducted on mechanical systems since. It was indicated by :;r. Cormly that he intends to conduct d e si gn revieus on unst tacchanical Design Class I systems; however, comprehensive design revieus, as required by QA Procedure No. ItE-6, have not yct been conducted within Machanical Engineering. Mr. Gormly stated that he could comply uith QA Procedurc "o. PRE-6, except for the requirement of co;r.- picting the revicu before the drawings are " Approved for Construction". Mechtnical Engineerinn, is, at the present tinj, avoiding thin requirement by marhing other entries such as " Approved for lestallation" in the change block on the drawingu. Mr. Lindblad had, prior to the intervicu, asked Mr. Coraily to develop a schedule of mechanical systems that will require design revicus. Mr. Cormly indicated that the schedule vould be fini shed June 1. The schedule was not available for the auditor on June 14, 1972. {_ivil Encincerine, Mr. Wollak, Project Civil Engineer, indicated that design revicus as required by QA Procedure No. PRE-6 have not been conducted within Civil Engineering. It ic the understanding within Civil Ens;ineering that since the conctruction drauings and the calculations arc donc by Civil Design-Drafting, the review and approval .f the construction drauings by the Responsible Engineer is the main design revicu. Houever, since the designers are under technical direction of the Responsible Engineer and a , l art;c amount of engineering vork has been done by the Responsibic Engineer

 ,       before the design is turned over to the decign squads, it appearn to bc questionabic whether independent design revicus are conducted. QA Proce-dure No. PRE-6 states that the design revicu vill be conducted by an engi-neer who had no significant involvement in the design under revicu. It further states that the revicuing engineer will prepare a rerort of his       '

work. Contrary to the above, reports of completed design revieuc have not been uritten by Responsible Civil Engineers. When the Civil design squad hac prepared the design calculations and construction drawings, the drawings and calculations are checked by a designer under the 3;uidance of hic immediate supervicor, who also super-vises the preliminary design. Since verbal instructions are conson between the Responsible Engineer and the designerc, and the instructions are not documented, it is questionable uhether the revicuing designer receives all of the necessary information which is essential for conducting independent / desii;n revieus uithout recourse to the Responsible Engineer. The check calculations, filed in folders in the draf ting room, do not provide a complete picture of the design revicu proccas, since the revicuing designer does not preparc a tcport of his revicu, as required by QA Procedure No. PRE-6. It is, therefore, difficult for a disinterested revieuer to verify whether the design una checked for applicable specific requirements. Documented evidence to verify that PSAR requirements, codes and standards, and assumptions, if any, are translated correctly into the construction drawings and specifications should be a part of the design revicu. M-2 h

Mr. E.11ollak indicated that he had been inutructed by Mr.11. J.

   . hinJblad to develop by June 1 a schedule of civil structures and couponents Q         which ulll require design reviews. The schedule was not available for the auditor on June 13, 1972.

Conclusion The audit revealed that design revieus, as required by QA Proc.e-dure No. Pnt-6, are generally not being conducted by the three engineering disciplines. Prior to the audit , Mr. U. J. Inndblad had instructed Mechant-cal Engineering and Civil Engiuocring to develop a schedule of structures and systems uhich will require design revicus. The design revieu schedules had not been developed on June 13, 1972. It appears to be the understanding uithin Civil Engineering-thet the Responsjhic Civil Engiucer's review and approval of the construction drawings constitutes a design revleu. It is, however, questionable whether independent design revicus are conducted, since the Responsible Civil Engi-neers are involved in the design. Design ruvicu reports, as required by QA Procedure No. PRE-6, have not been written by Responsible Civil Engineers. It is, therefore, impossible to verify whether the design was checked for applicable specific requiremcuts. Mechanical Engineering has not conducted design reviews during the last eighteen months. They are, at the present time, avoiding the requirement in QA Procedure No. PRE-6 by issuance of " Approved for Installh-tion" drawings instead of " Approved for Construction" drauings. Electrical Engineering had, prior to the audit, developed a system design review checklist which, when filled out, will hu used as a cover thcet for the design revicu reports. The chech calculations and the design-revicu checklists will, uhen filled out, comprise the design review reports and will then satisfy the requirements in QA Procedurc No. PRE-6. There appears to be confusion and a lack of understanding within the Engineering Department as to what a design review constitutes and hou comprehensive it should be.

                     " General Design Critoria for Nucicar Power Plants", l'OCFR50, Appendix 11, Section 3, Design Control, states in part:
                     " Measures shall be established to assure that applicable regula-tory requirements and the design basis, as defined in 50.2 and as specified in the license application for those struct.ures, systems, and components to which this Appendix applies, are cor-rectly translated into specifications, dravingu, procedures, and instructions.

Measures shall also be established for the selection and revicu for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems, and components. / Q' M-3

The desir,n control measuren shall provide for v ifyinti or chech-ing the adccpiacy of design, such as by the performance of desigu / revleus, by the tice of alternate or niup1ified methods, or by the performance of a suitable testing program. The verifying or checking process shall be performed by individ-unis or groups other than those who perforued the original design, but ul o buy be from the same organization." ANSL " Quality Assurance hequiremen'ts for the Design of nucleat Pouer Plants", Standard N45-3.11, states in part:

                 " Design revicu means the critical revicu of the desir;n in order to provide further as.surance that the actions leading to the design output such as drauings, calculations, analyn.is, and spect-fications, have been satisfactorily perfonned and the in forr.a t ion included in the design output is corrcet. '1here are inany ways of performing design verification, and variouc depths of verification may be required, depending upon the importance and comple:<ity of the design, the degree of standardization, the state-of-the-art, and the similarity with previ ".nly proven designs.

Regardless of the degree of standardization or similarity to pre-viously proven decians, the applicability of standardization or previously proven designs uith respect to meeting pertinent deci bn requirements shall be verified for each application. The results of design verJ fication ef forts shall be clear,12 docu-mented and filed and shall be auditable against the verification requirements identified by the responsible design organization.',' QA Procedur e No. PRE-6 states in part:

                 " Comprehensive revicus of Design Class I structures and systems shall be conducted to assure that the decign is complete and ade-quate....    . The Supervising Engineer vill designare an engineer to conduct the revicu. The design revicu uill be conducted by a qualified engineer who has had no significant involvemen t in the design under revicu. The revicuing engineer vill preparc a report of his uork. Design revieus shall consider as a uinimum the steps outlined in QA Procedure No. PRU-2, " Design Ucvelopment". When a design revieu uncovers an apparent deficiency in the design, the matter vill be referred to the nesponsible Engineer and the Lead Engineer for solution. The report of the design revieu will cover all such matters. The design revieu shall be conducted prior to issuance of " Approved for Construction" drawings.

M-4 h k

e A In order to coniply with QA Procedure No. PRE-6, there appears to be four main requiren.cnts which should be satisfied: T (V 1) Dacinn revicus shall be conducted on all Class I systans, structuren, and components. I

2) The{ revieu shall be conducted by individuals other than tho;;c uho perfortacd the original design.

3) The design revicti shall be clearly docutaented.

4) The design revicte chall be conducted prior to issuance of

                              " Approved for Construction" drauings.

It is the opinion of the auditors that the three eingineering disciplines are not adhering to the above requirci: tents set forth in QA Proceduro No. PIM-6 JENS ERLINCSSON Q*'<. a s d.' 'a 6 a , 1 *. --

                                                                 /,      .,     4- l

[ 60Tc'gl** VEf:NE L.. TEED (M IILM ' 6/19/72 b [ ( ) M-5

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O ATTACHMENT N PGSE AUDIT OF COMPREHENSIVE DESIGN REVIEWS DATED MAY 6, 1977 N l lO

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( g ATTACHMENT N s' PGWE ig FOR INTR A -COMPANY (D ES f No*f' omsson on *iktg RVD DEPAmmtNr JMcC ornnmtur QUALITYAgSURANCE GHA cr g en or L&frA ) e sus;rcr Comprehensive Design Reviews / 4VII MAY 2 5197'7 N WRJ Conformance to Procedure PRE-6 DYK c M "#j lf, BDt'd Audit No. 77016 acu u u ru DP Diablo Canyon Project d_EfffRAL Fltfr3 A NVWS ' [ May 16, 1977 MR. J. D. WORTHINGTON: Because the completion of the comprehensive design reviews for the Diablo Canyon Project is a prerequisite for fuel loading, an audit of their status was conducted between April 18 and April 29, 1977. The report of that audit is attached. Although the compre-hensive design reviews are nearly completed, some apparent design deficiencies which were identified during the design reviews have not been resolved. Also, several design review reports lack required approvals and the appropriateness of the proposed resolutions to a number of apparent design deficiencies appear to be open to question. These conclusions have been discussed with both the Diablo Canyon Project Engineer and the Vice President of Engineering. No formal corrective action is required at this time. However, fuel loading cannot begin until all of these de ign reviews are satis-( factorily completed and approve'd. M

                                                            . P. WISCHOW KGBouma(3692):eg cc RS3ain WHbarr GSBates JCCarroll PACrane, Jr.

JBHoch FFMautz/ RDRamsay l GVRichards ! J0Schuyler l CHSedam l BWShackleford l c MRTresler RWhite N-1 l l Attachment i

I Audib.o.: 77016 Irrzua D:ta: May 6, 1977 Page 1 of 7 PACIFIC CAS AND ELECTRIC COMPANY QUALITY ASSL*RANCE DEPARTMENT g

Title:

Audit of Comprehensive Denien Reviews Audited Organization: Diablo Canvon Project Auditor: Kent C. Bouma Dates Performed: April 18-29, 1977 1.0 Scope This audit was conducted to verify compliance with Quality Assurance Procedure PRE-6, " Comprehensive Design Reviews",and close out Audit Serial J1 76-8. 2.0 Conclusions and Exit Interview The designs of all the selected structures, systems and components have been reviewed. These reviews comply with the requirements of Procedure PRE-6 except that (1) some apparent design deficiencies identified by the reviews have not been resolved and (2) several design review reports lack required approvals. These activities are presently required to be complete prior to fuel loading. In addition, some of the proposed resolutions to apparent design deficiencies identified by the design reviews did not appear to address the question raised by the reviewer and in several instances the resolutions were not consistent with the FSAR. Because we are nearing receipt of the operating license for Unit I and the design work by the Engineering Department is likely to continue, it would be prudent to begin performing work, such as design verification, in accordance with the Quality Assurance Manual for Operating Nuclear Power Plants. This transition is addressed in Appendix F of this audit report. These conclusions and recommendations were presented to the Project Engineer on May 4, 1977 and the Vice President of Engineering on May 13, 1977. 3.0 Corrective Action Reports No Corrective Action Reports (CARS) have been issued as a result of this audit. The activities reported as incomplete and the resolutions considered open to question could not be considered as infractions or deficiencies until after the deadline for completion of the design reviews. A detailed listing of those activities considered incomplete and resolutions considered open to question is provided in Appendices C thro E. Performed by: W f*C ouma / N-2 Approved by:

                                                                    ._   [f               i R. W. MlfcKow l

   .                                                                                                                                                      dit Ns.: 77016 P;g2 % of 7

()

• Plan for Audit 77016 of Comprehennive Design Reviews Appendix A

1. Meet with Project Engineering to explain the purpose and plan for conducting the audit.

2. Obtain copics of the design review reports prepared by each Engineering Depart-ment.

3. Check the design review reports against the list of safety related structures, systems, and components selected by Project Engineering for comprehensive de-sign review.

4. Review design review reports not reviewed during previous audits for:

(a) consideration of appropriate requirements set forth in Quality Assurance Procedure PRE-2, " Design Development" (b) resolution of all apparent deficiencies in design by the appropriate Engineering Department (c) approval by the appropriate Engineering Department Chief

5. Convey results of the audit to Project Engineering at an exit interview.

 /\

C) N-3

( dit N?.:

                                                                    .          77016 P ga 3 ef 7 Appendix B Personnel Contacted                              g A preliminary audit meeting was held on April 18, 1977 with:

Mr. M. V. Williamson Project Licensing Engineer Mr. D. L. Polley Project Administrative Engineer Mr. K. G. Bouma Quality Assurance Auditor Personnel contacted during this audit were: Mr. E. P. Wollak Supervising Civil Engineer Mr. G. P. Brotherson HVAC Engineer Mr. S. Hanusiak Civil Engineer Mr. D. C. Landes Civil Engineer . Mr. E. G. Nichols HVAC Engineer Mr. H. J. Gormly Supervising Mechanical Engineer Mr. P. G. Antiochos Mechanical Engineer Mr. T. N. Crawford Mechanical (Instrumentation) Engineer Mr. J. W. Colwell Supervising Electrical Engineer Mr. R. A. Young Electrical Engineer Mr. F. J. Miller Electrical Engineer Mr. C. A. Partridge Engineer Exit Interviews were held on May 4, 1977 with: Mr. J. B. Hoch DCPP Project Engineer Mr. D. L. Polley Project Administrative Engineer Mr. K. G. Bouma Quality Assurance Auditor and on !by 13, 1977 Mr. F. F. Mautz Vice President of Engineering Mr. R. P. Wischow Director, Quality Assurance Mr. K. G. Bouma Quality Assurance Auditor N-4 h

                                 *                                               (                                     (

Audit No.: 77016 Pags 4 of 7 i

                "5                                                                          Append.'x C e                                                            -..

Civil Engineering Design Review Status

1. The design of the Turbine Building is incomplete because loading for the Component Heat Exchanger pipe supports is'not available.

2. The design of the Containment Structure rupture restraints is incomplete.

t-

3. The entry of the Civil Heating, Ventillation and Air Conditioning (HVAC) files into the Company file system is not complete.

A 4 The proposed resolutions applicable to the Auxiliary and Fuel Handling Building HV Control System do not clearly address all aspects of the apparent design deficiencies and a design procedure document does not , actually exist since the design was contracted out before the first issue of PRE-2. This is being coordinated with Electrical Engineering. Subse-

j. quent to the exit interviews, these resolutions were rewritten. They now appear to be correct and responsive to the apparent design deficiencies. ,_

i , i 5. None of the design review reports were approved prior to March 15, 1977 {. when PRE-6 was revised to require apptoval by the Engineering Chief. ] This approval authority may be delegated in writing. I. I { I r l IO N-5 l I

Aurlit Nr. : 77016 ( PI- 5 cf 7 l Appendix D Mechanical and Nuclear Encineering Desien Review Status g

1. Main Steam and Feedwater isometrics 445878 and 447119 have not been reviewed.

2.' The resolution of the seismic qualification of the Containment Fan Coolers is pending receipt of data from Westinghouse.

3. Resolution of apparent design deficiencies on the Containment Fan Coolers cannot be determined because the memos agreed to at the design review meeting have not been placed in the design review file (18.25).

4. Some resolutions to apparent design deficiencies on the Containment Fan Coolers accept quality records such as supplier manuals.and purchase documents with conflicting or incorrect data. Entry of such records into the Records Management System could compromise its effectiveness.

5. The resolution to the question of whether the Containment Purge Values are adequate for a postulated accident states that the question is outside the scope of the review.

Subsequent to the exit interviews, the Supervising Mechanical Engineer added documentation to the design review file explaining the technical reason no safety problem is felt to exist. This apparent design defi-ciency now appears to be prcperly resolved.

6. The reviewer's question of why a design criteria memo was not prepared for the Containment Purgc valves was answered by stating that such memos are only required for systems and these valves are part of the Containment Ventilation System. This resolution appears to be open to question.

7. All mechanical design revicu reports had been approved by the Supervising Mechanical Engineer prior to the revision of PRE-6 on ! Larch 15, 1977.

However, several of these reports were approved conditionally since all the apparent design deficiencies identified by these reviews had not been resolved. Design review reports not completed and approved prior to March 15, 1977 should be approved in accordance with the revised procedure when completed. N-6 g

                                                                                                                                     .                                                                           Ardit No. : 77016
                                                       '                                                                         -ar                                                                             P. J 6 of 7 Appendix E O

Electrical Engineering Design Review Status

1. No design review reports have been approved.*

2. Two proposed resolutions are pending receipt of data from suppliers and several imply that future actions are necessary before they are complete.

3. Some proposed resolutions do not specifically address the entire apparent design deficiency described in the report or the proposed resolutions deviate from Company commitments such as the FSAR.

4. Many of the proposed resolutions refer to reviews, analyses, tests and qualification records without indicating whether verifying documents exist or where they are filed.

)
                                                         *Since none of these design reviews were approved, specific recommendations were communicated directly to the Supervising Electrical Engineer, and only generic comments included in this report.

i D l l l N-7 1 l.

                                                               . . . . . . . . . . .....      j P        7 cf 7 Appendix F Transition between Design Verification under Procedure PRE-6 "Comorchensive Design Reviews" and OAP 1.1, " Design Develop-ment and Verification" Verification of design by comprehensive design reviews conducted to meet the requirements of PRE-6 should cover the design as it existed when the review was performed. These design reviews should be completed and filed.

If significant redesign occurs or design analysis is required for more adverse I conditions following performance of the original comprehensive design review, then the Engineering Chiefs must determine whether this additional design work must be verified. If verification is necessary it could be conducted in accor-dance with PRE-6 and reported as an amendment to the original design review report. Ilowever, design verification conducted within 90 days of obtaining an operating license should conform to the Quality Assurance Manual for Operating Nuclear Power Plants. The applicable procedure in this manual is QAP 1.1. This procedure requires the Project Engineer or equivalent design supervisor to determine what design activities require verification. It further states that the requirements for design verification are contained in ANSI N45.2.11. Section 6. While this ANSI Standard has explicit requirements, it would be very difficult to use directly without more specific Engineering Department implementing procedures. The Engineering Department should prepare for a transition to compliance with the Company's quality Assurance Manual for Operating Nuclear Power Plants by preparing procedures for design verification in accordance with ANSI N45.2.11, Section 6. N-8 h

O I

ATTACHMENT 0 l PG6E AUDIT OF DESIGN VERIFICATION DATED FEBRUARY 28, 1979 1 Q

P G e* E

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                                      , . . . . .                                             Le inn Veritication Diablo ranyon Auctt                        .o. 91320 N
         '*J           c                                                                       February 28 1379 II i'i           -

h g itk . .* . D. WORTillhCTON: g pJ An audit of design verification with emphasis on Hosgri seismic

                                                                                               <;ualification and fire protection was performed during January and q        7                                                                        February, 1979. The audit team included members from Engthetring i}uality Cont rul and Quality Assurance.

O e l Tlic attached audit report concludes that completed design verifi-cation conforms to requirements but that cuch of this work is o st ill in progress. Design verification is to.be completed prior to startup testing. Four discrepancies less significant than nonconformances were f identitled. They are: h { (1) Some atety-related equipment was inadvertently omitted from the review for qualification to Hosgri seismic criteria. {W (2) No verification of piping qualification to the Hosgri seismic criteria per the Engineering Department Manual was planned by clic piping section. (3) It is not clear to what extent quality assurance measures were implemented during qualification of NSSS equipment for Nosgri

         "),                                                                                                            seismic criteria.
         !                                                                                        (4) Lack of verification for changes to previourly verified design.
         >{                                                                                      Open item Reports have been issued to track the resolution of these discrepancies.

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                                                                                                      !* sue Date: 2/2H/19 Page     of 2                           l:_

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                                                                                                                                                .-J   '

J PACIFIC CAS AND ELECTRIC COMPANY QUALITY ASSURANCE DEPARTMENT  ;

      %                 ;ttle:                            Desicn Verification N               Aud ited os;;.snt zat ion /

q F.it 111t y : Eneineerine Department, Diablo Canyon Project Auditor : K. C. Bouma (Lead Auditor) E. Punzalan I bate, Pertormed: January 15 through February 22, 1979 7 1.0 Score

       *?

Determine whether: g (1) Previously issued Open Item Reports concerning design verification have been resolved and closed. (2) Structures, systems, and components that were given com-prehensive design reviews have had their qualification for the llosgri seismic criteria verified. (3) Reanalysis and design work to qual.'.fy the nuclear steam supply system (NSSS) for the Hosgri seismic criteria was verified. (4) Fire protection design has been verified. (5) Changes to previ,osly verified design have been verified

• and evaluated for effect on the overall design. .

2.0 Con,-lusions and Exit Interview Previously issued Open Item Reports (OIRs) concerning design verification have been resolved or are scheduled to be resolved during the qualification of tue affected equipment to the Hosgri seismic criteria. See Appendix A for deta.ls. The Itosgri Seismic Evaluation addendut to the Diablo Canyon FSAlt addrseses structures and systems or components of systems whose design was previously comprehensively reviewed. One exception was noted. OIR 038-79 identifies the inadvertent omission of the containment purge valves. Analyses and tests supporting tne Hoagri seismic evaluations are in progress. Verifica-tion of completed analyses and some resulting redesign is not yet complete. OIR 039-79 was issued to document the failure of the M&NE Piping Section to provide for verifying qualification of mechanical piping to Hos3ri seismic criteria per Engineering Department Procedure (EDP) 3.4 See Appendix B t'or details. 0-2

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{ . . 5 ',.L_lz' O, Audit No. 91120 P..g e 2 of 2 The eqii t pc:ent supplied under the NSSS contract is being seismically r . analyzed to the liongri criteria and necessary modifications are to bc designed by We t tngnouse under Purcnase Orders 4R-10592 and 4R-10600. = So i uality assurance requirerents were included in these puret.ase ordern. 3

                           *he IGandF. engineer administering these purchase orders han sent a letter rwiest iny, Westinghouse to identif y any quality assurance applied to this 9                      .ork.                            t> pen item Report 040-79 has been issued to monitor this situation.

7 Veritic.ition ut fire protection design, performed since issue of the Fire l'rutection Neview (FSAR Amendrent 51) a year and a half ago, has not been

    ~

aone. P ro,;ec t Engineering has now scheduled veritication of fire l g p ro t ec t ti,n des igt'. N utanees ti. previously verified design have not been verified and evaluated for ett'ect on the overall design. Open Item Report 041-79 has been issued 7 to resolve this discrepancy. See Appendix C for details. Much ot the Hosgri qualification work has been contracted to consultants

     --                    who are required to verify any analysis in accordance with their approved quality assurance, programs. Other than controls on preparation of contracts
     ")                    for such services, there are no requirements in the Enuneering Manual applicable to consultant design verification.

the lac k of procedural controls applicable to consultant design veri-tteation. along with other audit conclusions were presented at an exit interview held on February 22, 1979. Those in attendance were: Engineering Quality Assurance H. J. Gormly E. Punzalan K. C. Bouma i J. B. Hoch J. O. Schuyler T. G. de Uriarte , S. D. Krein E. P. Wollak D. L. Polley R. A. Young Performed by: , k

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                                                                                                                                           / . uq:A E. Puntal M               i Approved byt  7_

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                                                                                                                                       'W. A. Raymond Appendices A, H. C, D                                                                       0-3
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_ 'r-$ x :: .. . f Audit Nu. 9137; Appendix A Page i of I 3 1 APPENDIX A PREVIOUSLY ISSUED OPEN ITEM REPORTS (011ts) I APPLICABLF. TO DESIGN VERIFICATION h 3

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                          . OIR 002 Issued 2/26/76 Assuned to:                           Supervising Civil Engineer, Diablo Canyon Pro,'ect                                                                            .

3

Subject:

The Component Cooling Water (CCW) Heat Exchanger pipe support loadings have not been reviewed. N Resolution: The required review will be performed as part of the Hosgri reanalysis of the Turbine Building.

2. OIR 003-78 -

Issued 2/28/78 9 Assigned,to: Supervising H&NE Engineer, Diablo Canyon Project

Subject:

The seismic qualification of the Containment Fan Coolers has. O not been performed and documented. Resolution: Perform and document the seismic qualification as part of the liosgri reanalysis.

3. OIR 010-78 -

Issued 2/28/78 Assigned to: Supervising M&NE Engineer, Diablo Canyon Project

Subject:

The potential effects of fluid line failures inside contain-ment on nearby electrical penetrations is undetermined. Resoluticn: Potential adverse effects are covered in Section 3.6 of tha FSAR and by LOCA qualification. This is now clo, sed. 4 OIR 047 Issued 3/28/78 Assigned to: Lead Electrical Engineer, Diablo Canyon Project

Subject:

Seismic qualification test procedures used by Wyle Laboratories do not contain acceptance / malfunction criteria. desolution: The test criteria have been added to the addendum to the FSAR for the Hosgri Seismic Evaluation. This is new :losed. 0-4

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    ~.                                                                                                                                                                                                                    Appandix !s Page 1 of 4 s

APPENDIX B N STATUS OF HOSCRI CRITERIA DESICh VERIFICATION t T i 6 l .o Att.achments ' ' to this appendix identify the means and st.itus

                  -                                                           s f verification of the qualification of st ructures, systers, and components to Hosgri celsmic criteria for each engineer-l                3                                                         ing discipline. An assessment of conformance to Engineering l

Department Procedure (EDP) 3./ , " Design Verification," is 3 included. O . Attachment 1 - Civil Engineering g Attachment 2 - Electrical Engineering C Attachment 3 - Mechanical and Nuclear Engineering l o Quali11 cat 16n may be by analysis or by testing. A separate i i veritication activity is not necessary when the qualification sethod used is testing to the most adverse condition. How-ever, when qualification is shown by analytic metheds, it must be subsequently verified independently by a design review or alternate calculation. r I P i L i e e b I

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[ > Audit .s o . 91 M0 AppendIz N Page 2 ut 4 Attachment 1 M Status nf Civil Engineering Hosgri Design Verification Q Livil st ructures, systems, and components are being qualified to Hosgri seismic

                      .:rs teri , by analysis which is reviewed for verification. Verification by consult-a n t '. is controlled by their approved quality assurance program. The consultants I              uwd are URS/blume (URS), EDS Nuclear (EDS), and Harding-Lawson Associates (Hl.A).

T he stea n s and status of design verification is tabulated below: N Structures Method Status

     .3 Containment                              Review In-House                                    Completed 1/19/79 O                         Auxiliary Building                       Review In-House                                    Completed 12/13/78 Turbine Building                          Review by URS                                     Due 4/1/79 O                          Intake                                   Review In-lloose                                  Due 3/15/79 Ventilation                               Review by EDS                                     Due 3/15/79 Tank Outdoor                                  Review by URS                                      Due 2/28/79 Diesel Fuel 011                          Review by Hl.A(I)                                  Completed 1/3/79 Min.cellaneous Auxiliary Saltwater Piping               Review by Hl.A(I)                                  Completed 12/13/78 Spent Fuel Racks                         Keview by EDS(I)                                   Completed 8/31/78 Cranes                                   Review by URS                                      Due 4/1/79 All Civil structures, systems, and components given previous design reviews are included in the above table. Both completed in-house reviews were checked and found to conform to EDP 3.4 (I) The cover sheet shown on Appendix B to EDP 3.4 has been attached to completed consultant verification reports to show PGandE acceptance.

0-6 h l l

                                                                                                                                                           .a.

Audtt No. 9I320 Appendia b I-PJ4e I 01 4 h AttJChment 2 ' Status o t' Electrical Enyineering husgri Des ign Verit icat ion 7 O F.lectrical equipment and Class IE instrumentacion was qualified to the Hongri

     ~
                    ,e i .rr i c criteria individually or by type rather than by system. This makes it 1 pra.tical to tabulate status by system for comparison with the original desig,n g         reviews.t!) Instead the status can be presented by grouping electrical equipment
                   .ind Class 1E instrueentation by the qualification method used. These include:

N g 1. Testing by Wvle t.aboratories(2) '

       ~                             This testing, conducted under PCandE Contract 5-61-77 is now complete. Reports have been received f or all but the last a                              series of tests. Except for lack of a specific cover sheet, these reports, augmented by Section 10.3 of the Hosgri Seinmic C3                              Evaluation

• addendum to the FSAR, conform t'o EDP 3.4

                               . ' . Analysis by Civil Engineerinc Seismic Section This analysis is complete but has not been independently ve r i f ied. No schedule for completion of the verification has                                            j been established.                                                                                           '

3. Acceptance of Westinghouse WCAP Reports These reports contain analysis and test results showing seismic qualification similar to that required by EDP 3.4 They were prepared by Westinghouse, and PCandE checked the seismic qualification levels against Hosgri seismic criteria.

4 Tests Conducted by Instrument Suppliers Qualification levels in supplier provided test reports were checked by PCandE against Hosgri seismic criteria. (I) A check against original system design reviews did not tavcal any equipment overic,oked, but at the exii interview it was noted that Heat Tracing had not been qualified. Electrical Engineering has concluded that the heat tracing system need not be qualified to the Hosgri seismic criteria to assure safety of the public. (2) Sec quality Assurance Audit Report 78006, dated April 7, 1978. 0-7

             ,,   , ;y,    99       g-              ;c  r-- ~ ~ g y ssa_m g +s -                -

l [ .h, ,g 9

       ?,       *
    .1 Audit No. 91329

( Appenst i m s < f P.i r e 4 of 4 W t

     )                                                        Attachment 3                                                         ,

l

          *o                                 Status of :lechanical and Nuclear Engineering
    )      .

a m Losgri Design Verification i i 'O Mech in i c.s l items are being qualified to the Posgri seismic h _ criteria individually or by type rather than by syster. l 0 This makes it impractical to tabulate status by system for

   )

1 comparison with the original design reviews.(I) i f  ?! Qualificat ion of mechanical equipment was performed I i .,

            '                     in-house by the Mechanical and Nuclear Engineerir.g (M&NE)

Department, primarily by analysis, and is essentially com-

          -)                     plete. Westinghouse has agreed to p ovide an engineer to verify the qualifications under Purchase Orders 4R-10592 C3 and 4R-10600. The work is to be perf armed during March C3                     1979 under the supervision of the Project Supervising M&N Engineer in accordance with EDP 3.4 Mechanical piping is being qualified for the Hnsgri seismic criteria by computer analysis of the pipe stresses followed by evaluation and, where required, redesign of hangers for resulting loads. Some analysis is performed by consultants who verify their work in accordance with their approved quality assurance program. In-house analysis and redesign were not scheduled for verification by the M&NE Piping Sec-tion in accordance with EDP 3.4         OIR 039-79 was issued to resolve this discrepancy. At the exit interview, the DCPP Project Supervising M&N Engineer stated he was having Design j

1 Draf ting perform the required verification. OIR 039-79 will monitor completion of this work. 1 Mechanical instrumentation is included in the remarks concerning Class IE instrumentation on Attachment 2. (1) A check against original system design reviews revealed that the containment purge valves had inadvertently not been qualified to Hosgri seismic criteri1 This is being corrected. Radwaste nystems were intentionally not verified for Hosgri. 0-8 1 i i

               .P'       E Audtt No. 91 L'il 1

Appendix C l Page i e,I I d

      *3
      *E e

d k APPENDIX C g

                .o                                                           \'ERIFICATION OF DESIGN CllANCES
                * *n 2

O Those design changes documented on Engineering Change Order

       ;                                           (ECO) forcs and their replacement, Design Change Notice (DCN) f                                           forms, were checked. With only minor exceptions, these forms j

had been properly completed and approved. 1 1 i Each form has a specific block where the initiating engineer l M checks whether the change affects design verification. The ll' l _ vast 2jority are marked No; however, two DCNs are marked "l e s . These are:

       '                                                                                  DCO-E-C-032 l

1 O DCO-G-E-500 O No verit ication of design changes documented on ECOs 'or DCNs has been performed and none is planned. The tunction of the check-off block on the DCN form is unclear and its use is not consistent with the requirement of the Engineering Manual Procedure 3.4, that all changes to previously verified design must be verified. Open Item Report 041-79 has been issued to resolve this discrepancy. i O o-9 l 1

O ATTACHMENT P PG6E FILE REVIEN RELATED TO DESIGN CONTROL DATED OCTOBER 6, 1981 1, e t i O

   .

• A T T A C H M Eli.T. P .

     + . . . . . . . p.sv. .im tG&

P G r.=>E i FOR INTR A-COMP ANY USES I From Division or Department QUALITY ASSURANCE

                                                            .j Fitt No.                                              b'    g RC LETTem or susaccr                                                  .[

To Division or Design Control g

                                                                ,I Department                                                                                                 i October 6, 1981 MR. W. A. RAYMOND:

Per your request, we have reviewed our files for Quality Assurance involvement within the scope of Design Control in order to respond to some questions regarding Quality Assurance activities in this area. We have listed on the attached sheets five questions that have been posed, documentory evidence that applies, and a short response to each question. We hope this information is useful. D. S. AARON A h'M 7 GU/DSA:eh Attachment I P-1

I. Does the record show that Quality Assurance spent time auditing Engineer to assure that a system was in place for design developments, transmission of design data, and design verification? A. Program Audits: There were a total of 37 program audits concerning design control and related activities performed on Engineering for their compliance to the procedures in the Red (construction) manual from 1970 through 1976 (the Yellow (Engineering) manual was issued in 1977). These audits covered the following areas: 1970 DC Program Audits

1. Civil Engineering

2. Civil Engineering Computations

3. Main Stream Isolation Valves (Spec. 8721)

4. Electrical Penetrations (Spec. 8806) 1971 DC Program Audits

         &5.        Specification Preparation, Review, and Release X Y6.      Drawing Preparation, Review, and Approval 7 M 7.       Civil Design Computation
         )( 9 3. Engineering Department Component System and Structural Design 1972 DC Program Audits

9. F6D Hydraulic Support Struts (Spec. 8794)

10. Station Battery Chargers and Electrical Storage Batteries (Spec. 880C

         . Ril. Comprehensive Design Reviews

12. Engineering Releases y f-13. Comprehensive Design Reviews 1973 DC Program Audits

14. Mechanical Control Boards (Spec. 0725)

15. Supplier Inspections

            .-16. Design Development

17. Engineering Department g:18. Comprehensive Design Reviews P-2 h

 =

• i l

Shast 2 cf 9 1974 DC Program Audits

19. Emergency DC Lighting System N '20. Engineering Material Memos

               .21. Engineering Department, Seismic Requirements of (Spec. 8771) 1975 DC Program Audits

45. Electrial System Design Reviews

23. Mechanical System Comprehensive Design Reviews

24. Purchase Document Change Control

25. Design Review of Civil Systems X/ 26. Design Review of Electrical Systems X >'27. Supplier Bid Review and Award

         .)( P 2 8. Design Review of Civil Systems and Structures g}'29. Electrical Syste[n Comprehensive Design Reviews M30. Engineering Department

31. Purchase Document Change Control

             # 32. Engineering Releases
             /J3', Mechanical System Design Reviews 1976 DC Program Audits

34. Civil Engineering Design Reviews

35. Safety Analysis Report Changes

          .5p36. Electrical Wire and Cable Installation
              .37. Engineering Design Reviews B. Ten Specific Quality Assurance Audits of Design and Design Review or Verification:

While there are many audits of specific specifications in the files, we have selected the following sample of pertinent audits on Design for an examination of content:

1. " Audit of Quality Assurance Procedure No. PRE-2 (Design Developmer;t)"

performed June 20-22, 1973 by B. G. Good of QA and J. Comes of Civil /Architechtural. Areas of concern that were identified and transmitted to Engineering included:

a. No formal design procedure.

b. No formal checklist. data sheet, or other form delineating design requirements.

P-3

Sh;st 3 cf 9

2. Audit of Seismic Analysis of Equipment in Specification 8771 (Control Room Complex) performed December 16, 1974 by T. G. g de Uriarte of QA. The audit concluded that all required W reviews and approvals had been made.

3. Audit of Engineering Material Memorandums performed during t.;

month of December 1974 by K. G. Bouma of QA. The audit identi-fled a few EMMs that lacked QA requirements and that a procedural modification was needed to allow QA to monitor EMMs. Attached to the audit is a letter frca the audited responsible engineer, T. N. Crawford, taking exception to much of the audit.

4. Audit of Electrical System Comprehensive Design Reviews perfor=ed during the month of February 1975 by K. G. Bou=a of QA. Areas of concern that were ident'ified and brought to the attention of Engineering included:

a. No documented reason for omitting various aspects of the design from the design review.

b. Potential design deficiencies identified by EDS Nuclear had not been resolved.

c. Reports did not contain required approvals.

5. Audit of Mechanical System Comprehensive Design Reviews performed during the months of April and May 1975 by K. G. Bouma of QA.

This audit also identified the concern that there were no docu-mented reasons for omitting various aspects of the design from the design review.

6. Audit No. 75-6 of Mechanical System Design Reviews performed during August and September 1975 by K. G. Bouma of QA. This audit was a follow-up to verify resolution of findings in a previous audit. The audit identified two areas of concern that had not been resolved:

a. Failure to justify design aspects not considered by the review.

b. Lack of required supervisory approvals on design review reports.

7. Audit No. 75-9 of Civil Engineering Design Reviews performed during the month of September 1975 by K. G. Bouma of QA. The audit identified the same deficiencies listed in No. 6. above.

8. Audit No. 76-6 of Civil Engineering Design Reviews performed during the month of February 1976 by K. G. Bouma of QA. The concern from this audit are statements by one of the audited engineers (M. V. Williamson) which are :*_tached to the audit.

P-4 h

Shsst 4 of 9

8. (continued)

{/} The engineer makes the same statement on three attached sheets of commenting on Design Review of Spent Fuel Racks, the l Auxiliary Building, and the Turbine Building. The statements read:

                           "The seismic inputs to the design were developed by John Blume and Associates, a consultant. The design review initially omitted consideration of the adecuacy of the Blume work. The work of a qualified consultant of high reoutation in a spe-cialized field such as seismic and structural analysis was, and is, considered to be outside the procer scope of this design review."

If this was the approach taken by the civil engineers on all Blume work and also the mechanical engineers (for hangers), then we would have never found that the inputs were wrong because of the flip-flop of the unit sketch. If we did not question the seismic inputs nor how Blume arrived at them, how could we ever discover that Blume mistakenly calculated them?

9. Audit No. 7 6-8 of Engineering Design Reviews performed during the months of July, August, and September 1976 by K. G. Bouma of QA.

The audit . was perfor=ed to verify if reviews conformed to require-ments of PRE-6. No significant findings were identified.

10. Audit No. 91320 of Design Verification performed during the months of January and February 1979 by K. G. Bouma of QA and E. Punzalan of EQC. Areas of concern identified and brought to the attention of Engineering included:

a. Some equipment was omitted from the review for qualification to Hosgri seismic criteria without justification.

b. No verification of piping qualification to the Hosgri seismic criteria was planned by the piping section.

The record shows that Quality Assurance spent a great deal of time auditing Engineering to assure that a system was in place for Design Development and Design Verification. There is little evidence, however, of any audits that focused on the transmission of Design Data until R. J. Pillers' audit of URS/Blume in 1977 (see page 1 of D. S. Aaron's letter of October 5, 1981 on the URS/Blume file). Another area of con-cern has to be the lack of a documented comment by QA on the statements by M. V. Williamson avowing that a review of design inputs for work by a qualified consultant (Blume) is not necessary. l l l O V P-5

Shtcst 5 of 9 II. Did EQC audit the Design Verification / Design Change areas after the EQC manual was issued? A. Ten audits of Design or Design Changes perfor=ed by EQC since the issuance of the EQC manual are listed as follows:

1. Audit No. EQ 7801 dated May 11, 1978

2. Audit No. EQ 7806 dated November 17, 1978

3. Audit No. EQ 7807 dated January 12, 1979

4. Audit No. EQ 7902 dated April 3, 1979

5. Audit No. EQ 7907 dated August 2, 1979

6. Audit No. EQ 7911 dated August S. 1979

7. Audit No. EQ EDS dated February 13, 1980

8. Audit No. EQ 8002 dated February 25, 1980

9. Audit No. EQ 8008 dated June 26, 1980

10. Audit No. EQ 8013 dated January 22, 1981 EQC has conducted a number of audits in the area of Design Verification /

Design Change since the EQC manual was issued. Audit No. EQ 7806 dated November 17, 1978 was conducted to complete the corrective action for Nonconformance Report No. DCO-78-EG-010 issued after the Blume findings by R. J. P111ers in November of 1977. Audit No. EQ 7806 found two more supoliers (out of 25 currently active at that time) that were doing design activities other than those activities they had been qualified to do. P-6 g

Shsst 6 of 9 I III. Af ter becoming aware of URS/Blume, did QA and/or EQC follow-up on other consultant contracts that involved seismic or that might have been work-

 ' ( )-

ing on contracts without proper QA requirements in the contract? A. Our files contain two sheets listing the 13 consultants that are involved in the above-mentioned areas that have been used.by PGandE. These two sheets were prepared by R. J. Pillers after his audit of URS/Blume identified this . requirement. A review of our files reveals that all-13 were reviewed and qualified per our requirements except Science Application who completed their work for PCandE at the time R. J. Pillers investigated the matter.

;                            The following is a list of the above-mentioned consultants:

SUPPLIER LATEST DATE QUALIFIED

1. Stewart Smith June 2, 1978 i

, 2. Wyle Labs (Norco) November 29, 1978

3. Wyle Labs (Huntsville)

July 13, 1976

4. NUS July L7, 1980 (EQC Audit 8001)

5. H. B. Seed July 22, 1980 1 6. Tera December 17, 1980

7. NSC (Quadrex) May.30, 1980 4

;                             8. Harding-Lawson                . April 25, 1978

9. Dr. B.= Bolt June 2, 1978 l 10. Dr. C. A. Cornell June 2, 1978

11. EES July 27, 1979

! 12. URS/Blume November 2, 1980

13. Science Applications NO FILE i.

i A review of our files. substantiates the corrective action required on Nonconformance Report DCO-78-EG-010 identified by R. J. Pillers' audit of URS/Blume. The NCR, which was verified as being resolved i by EQC Audit No. EQ7806, was issued to initiate a review of all

other active consultant contracts for similar trouble. QA and/or QC has followed-up on all other consultant contracts for problems identified in the Blume audits.

IV. Did we review all past contracts to assure that all consultants working on safety-related contracts have been reviewed for adequacy of QA programs? ? A. An investigation of a computer printout of DCPP contracts reveals l - that all the.following pertinent contracts have been re iewed: P-7

                                                                                                       )

l L

Shast 7 of 9 CONTRACT RESPONSIBLE DISCIPLISE 5-86-79 Civil 5-73-79 Mechanical 5-12-80 Civil 5-60-79 Civil 5-70-79 Civil 5-29-80 Civil 5-16-77 Mechanical 5-5-75 Projects Blume Civil 5-25-72 Civil 5-58-77 Civil 5-1-79 Civil 5-38-78 Mechanical 5-43-78 Civil VP&GM 72240 Civil VP&GM 71804-A Civil VP&GM 71783 Civil 11-68 Civil 5-13-75 Civil 5-59-77 Civil 5-51-74 Civil 5-68-77 Civil VP&GM 72849 Civil VP&GM 70539 Civil VP&GM 64376 Civil 5-31-69 Civil 5-9-70 Civil 5-19-70 Civil 5-20-70 civil 5-9-75 Civil 5-1-72 Civil VP&GM 70163 Civil 5-8-73 Civil 5-17-73 Civil 5-22-75 Engineering Research P-8 & W

Shast 8 of 9 CONTRACT RESPONSIBLE DISCIPLINE () 5-86-79 Civil

                              .5-8-75                    Civil 5-60-73                   Civil 5-8-74'                   Civil 5-30-/4                   Civil-5-42-74                   Civil 3-67                      Civil 4-67                      Civil 10-68                     Civil VP&GM 64845                   Civil VP&GM 64481                    Civil
                             .5-73-80                    Civil

, 5-73-79 Mechanical-

                              '11-68                    Civil 5-61-80                  Civil VP&GM 71783-C                 Civil 5-48-78                  Civil 5-58-77                  Civil 5-3-81                   Civil 5-16-81                  Civil

{ 5-20-81 Civil 5-23-81 Civil All 57 contracts involved show a review of the consultant's QA program has been done. l i V. Did QA give training sessions for indoctrination of Company personnel in QA Policy and Procedures? A. Prior to 1978, training was given individually by QA personnel while assisting Engineers to do their tasks. That was found to be quite j effective during the formative years of the program. In 1978, QA ' started a program of formal training sessions as represented by the following four records of training sessions-given in 1978:

1. Training session given at DCPP to key plant personnel on the Policy Section of the QA Manual for Operating Nuclear Plants -

R. P. Wischow, Instructor. ! p P-9 i - 4 I-

Shiit 9 of 9

2. Training session given at the General Office to key Engineering Department personnel on the same subject as above -

R. P. Wischow, Instructor.

3. Training session given at HBPP, Unit 3 for key plant personnel on the same subject as above - R. P. Wischow, Instructor.

4. Training session given at Department of Engineering Research in San Ramon to key personnel on the same subject as above -

R. P. Wischow and B. G. Good, Instructors. Upon issuance of the Engineering Quality Control Manual, training sessions in implementation of procedures has been an ongoing activity. Annual audits of training indicates that all quality control groups in PGandE currently have active training in procedure implementation. l P-10 g

7 O ATTACHMENT Q PIPE SUPPORT AND RESTRAINT DESIGN GUIDELINES IN 1976 i O 'l

    -- - ---m, .._. - -  n --, ---- - -      --,    ---.   --

\ rage 4

/^T
\    t V               2. An increase of 80 percent of minim a yield at room temperature during hydrost atic testing, but not exceeding 25,000 psi for st cel of unhncun physiccl properties.

3. An inercase of 90 perecnt of minicut- yield is alloued for double decir,n enrthquehc.

The allcr.': Ale stresa for dif ferent lo'sd;.nc con:!itiev arc rut:.-:rized in Table 2. TARTE 2 - A li c:r:! l e S t rer..m-

                                                                    *Alle.cnb!c strc sc fer A3f, Support ina, Component         Allo.:nble         St ruc ture Steel M:tcri -)

1.cac's Strcnn Lt.1) Tegg tt're /'. f. Q ' + Til, DL S;g 12.6 Til, DL, TV , nVGT , Dr 1.2 Sn 15.0 111, DL .8 S y 28.8 THA, DL, TV, nVOT, Di>C .9 Sy 32.0

• Use 50 1,orce.it er the elle et.bic strecs for A36 plate n tcrial t all temper atures i htn load.-d in the E directic.n (I .c. , tensic. in ;he plane of the ti ichner.s of the ro.tcrial). This does not F.pply :o other pinte n2terial er A35 structural shapes.

Uhere: DL - Pipe Deadload DC - Design Earthc tal:c DDE - Dc.uble Design Ecrthquake FV - Fast Valve Clocure Load RVOT - Relief V:lve - Open System (Tra ns ien t ) Til - Therna1 Expaucien Leeds aL Norca1 Operatint Tc: perature TilA - 1heran1 D.lansion Loadn from Accident Ccnd i t. ion s llT - 11ydroc1.a tic ') est Lond 51 - Allcr. nble St resu V::lae at Cpera t ir.r, Tempera ture 1 Sy - Mini:num Yield Strength REFr.ME!:rP.S_ 1 Manual of Steel Construct ion, Sixth Editica, AISC, 1957 9

         ~ Manual of St eel Construct ion, Seventh Editica, AISC, 19 7 0, P. 5 -4 '.

(] ' Q-6

_- ,- u ., _ , s .,~ ,g. -. ,

                                                '- /         h,'      , !_                                           ff

, }T.C"J.!!ICAL (a ITJCI.17.R E .Cl::"TRII:0 146.10 rnd 146.20 Guidelines for Decie.n of :'ig 2 Su ppor t:.: ani 1 cutrai: tr. 7 h Di:.blo canycn Site - Unitt 1 cr.d 2 Augur,t 16, 1976 1D1. I. T. 1%LL: The attr-ci:cd : ciddit.28 for den!;.n of r.u;; rtt h:'v e 1 c':r 6 l e' :.J fron re G.cr .::.er. li: te 1, cc:;o ;, m:d 'S t. c:c' 7;..." fo: t! . Sinbl o C:a * , e.. u . i .< . Uhil e ce .c yc ! e e r. ar i n,.d :.; th . . i'i o.:r de t.4,.-

                                         .                                             t !. .; . cre :..ver f or .: H c: . ;

the int.cntior. cf t!. .. let t.ar i: to do thic.

                  /. t n Irtcr dnt:., a dre cing 1:: y t o i r.:                      .J rich < en        1.i. ~ ir ti:: t ocm . inf c r :;tirr on it if it ic felt nuce.tr; f b-                                c. li i n n e h  .
                                                                /i . G . t .,n ! . . .a.

AC.' (2.lO?)/vje At t:.1 : :nt cc: 1?J ain.ib in d/ IIJ G c.rn:' y

         /.14h u n REl'a ci cr Q-2                                                                        g

, y, (v) The following r,eneral guidelines are innued to provide desir.n criteria for piling supports and rentraints for the D.Mblo Canyon Poccr 1*lart i Project. For subjects chich are not covered in thin decument, the decir.ticr should refer to AIM 1 B31.1, U31.7, and the 'iss r.tnn:'crdc SP-5G and SP 09 ter guidance, ac ucil as the catalens and manuals publi shed by component m3 nu f a c t'!re rS . I. I'r.iaY1"'r W'irr: m S

1. All rntericls shall be of ucidabic quality and shn11 be of the net.s type er of co pntibic ccq onition with the roterial to which it is to be ucided sn:1 of cor.parabic yiele and tensile strent,th or higher.

2. The desi;,n a llev.,i.le strc .c en fillet UM d.c for suppc.rt ia:r.burc it. 13,600 psi.I for '.-ze bhed a t tact -

nents to the pir: , th: allo. hle 1ccdr. p.iv:.n in ITI Standt.rd ES-26 chould be uced.

3. Attachunuts welded to PGLE Code Cle: A piping sbill r be welded ei th ccaplete pen:i ratinn ucida coi,t <,ut p: r Figure 1-727.4.4(a), B31.7.(7)

4. In joints connected only by fi]Iet veldt , the ninit "-

size of fillet teld uccd is ac rhe.:n in Table 1. L'e ld size ic d.:.terstied by the thichnncc of the tro 1. ort" joim d, except that the ve id size need not exceed the thichneta of the thimier pcrt jeined uple.tc a it.rger size is required by calculated strer,s.2 _TA P. I ." l 11aterial Thichucss of liinirstn Sire of Thi ci:er P. r t Joitied Pillet 1' eld To 1/4" Inclusive 1/P" Over 1/4" to 1/2" 3/16" Over 1/2" t o 3/4" 1/4" over 3/4" to l'c" 5/16" Ov e r l '. " t o 2 '," 3/S" Ove r 2 '," t o 6" 1/2" over 6" 5/8" p) t Q-3

  \

\ Page 2

5. All field uelds made on beacc and girder under load or columns chall be parallel to the longitudinal ancs h

of these combers. i

6. Uhenever possible, velding to t.he pipo shall be avoided.

II. G.*i P E T Q The maximu:a total gap allvacd in the t.cienically rcutrained direction is 1/8". In ncn-rcs trained dirceti one. , t.ha su r par t dccign chall allou clearcnce for thermni and scicric move: cent of'the pipe durint. nor:ac1 operatia;; or accident conlitions, as vell as an insta11cticn tolerance of +1/16". 111. ALT.0;'A h!" prrT ECT10': I!! lir.STp.#.I?TTE For a seiccic Clt.rs I rectraint, t he static deficct Ion of t he restrcint in any rcccraint direct ien, due to the vei;;ht of the restraint, pipe, yl l'e insula tica , fluid , and pi pe a ttachmer4, tras t not c>:cced . 0.15" . For non-c .:istic Clarc 1 rec trr ints , a unnitc.uu deficction of 1/16" is allowed. IV. ACerprA",Lt tv,;'.; T10" op suppc:1T te<'! Tio" The denign loc.-tion of the supports in a rtraight ren of pip:: may deviate by 11 0.D. of the supparted pipt from the theoretien1 location provided it is not adjacent to equirc nt nor:.lcs or hc asy valves, in which crue, prior approval frer the Project Engiin er chall be required. If this relocation tolerance is utilized by the designar, he should specify on the banper details the revit.ed relocation tolerances for the hanger in:,tsllations. V. T ESTRA 1; r towg Aim pIspl c->:cT op 7pt p1p, The forecc, morents, and pipe displacer cut data for hanger desig:" shall be obtained from the therral, seic:r.ic, and deadload co:cputer output for ena ly::ed pipini, systeraa. lhe design scicnic Icad cf Clacr 1 pipe supports anl rectraints for piping not :.c ? smica ll y analyzed in tuice the cG Lit:cted combintd weight of the piping, fluid (including hydrotest fluid), innulation, and piping at tachment acting downuard and in the horizontal di rect ion. All restraints shall be.desi$ned to withstand the static ond kineatic frict'Pdn dueFor to rela t ive i.ievewnt of the pipe u; t.h res pec t to the restraint. piping systen not theren11y analyned, t be hanger desit.ners shall a lua ys provide clearance for the anticipated therral enpansien oi the piping system. Q-4

a r n Page 3 l VI. LPRU*CS, SmfST,T.RS. AW OTH 1 PI'P.CP/.SPD SUppmT COMPC;FTIS The allouabic load recemended by t he inanuf: cturers of the cprin: s,- snubberc, and ot.hcr Inrchased cupport coraponents shall be used fer pipe hanger designs. Hanger rods shall be subjected t.o tensile Ic:< ding only. Rod han er assetr.blice shc11'be designed to allo.7 c.ntici pate d ther:n 1 hc.rh:s nt r l mavere?nt without subjecting the pipe to c:ctran:rous leeds. T'.<> : e n ir s r swing angle due_ to hori::cata.1 pipe travement cbould bc less LI.r.n 4*, If the cuing cugle of the rod in in c::ce::n of 4* nnd/or the totc1 movertent ic in cr.eces of tuo int.hes, the 1. ant,er shall be offset two-thirds of the ther:r: l raovce::nt agi'r.st. t.he (irection of' move:nnt. The snabber assembly sh.,11 be of f n<.t two-thirds of the _ther 21. nover ent in the cc,1d positicn if the cuing eng1: cnccede 3* and/cr tha tot:1 narce. nt of the point of attt.chc cne on t he pipe is in excess of Luo inches. The n!.dpoint of ther:n1 travel f er cprin",r- nud : nubber strchs c.h:11 he set as tha tridpcint of the total t.ruv.:1 ; .th het a:-d ccal setti ;,c ectabliched ecordin;;1. j VII. CNnrTE r/.Sm?m' The allc.:abic lands and decign procedures for cre. bedded concre.te fastennrs t.rd cnpansion holtz, shall be in confurre. nec ui th f0:1 Standards 049324 and 054162. VIII. A LLO' .*A ",i.7 STP.f'ES F07 SU"N"!T STpVCTUn '.L CN2r0:5?NTS The allouchle stress for t.he base noterial of all parts of supportin3 and rectraint ascer.blies shall not enceed the cppi epricto S ; 1vc lue

                   - taken f ror.1 the A1:31 D31.1 Allouable ~ Stres:. Tcblev at t h e r.an I r. ur.1
                   - operr. ting ter porature of the part in cucat ion.       Desi gn tc:.p:r..ttees for parts of supporte end rectrnints in direct centact wit!. pip 2                 ..ha
                                                                                                        <    ll be the mximura op rnting tenparature of the' pipe.             Parts of hany rr.

and supt.crtn not in direct cont act eith the -pipc cn ' ext.:rior to 4.ny inculction ttay be- der.igurd lor one-third of the re:ni:.u:' c;' erat.w, . tempercture or arabient tecipera ture, ubichever is gre. iter. An increase in allouabic stress chall he permittcd as follo.:n:

1. An increase of 20 Percent for Short Tii..c Overloading -

( 4.One Percent of Operating Period), t , / - Q-s

Page 5 g. 3 American Itat icani r.tandard Code for Po.rer Pipinc., A::"I S31.1, 1973 g 4 Uelded LccJ E:cring At tachn: nts to Prescure Rctainint; Piping I'ateric1r., PFI Sta:.'ard ES-26, July 1973 5 MSS St ar.dard Practice SP-50,1975 6 liSS Stendard Practice SP-59,1966 7 USA Standcrd Code for Pre'.,s ure Piping,1:ucle.ar Pocer Piring P31,7,1909 Q-7

                                                                                                              /8- /7-7[/                                '

7-[ 7h ATTACHMENT Q

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                       ----2,--DE0!C'! 05.000 TOR
                                                                                                                       -*'5C ::AT -C0 LT TO 20, ' 00 TOI,                                                                                     .~

CEI?AIC 'C!l' IE-I'ER IC 22iwN 22,000-20!. .

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  '                                                                                                                                                                            s O.                                                                                                                                                                                                      - f,.

YOR A SEISMIC CLASS I RESTRAINT, TE STATIC NEW CONS . "CTIO3' . SE d<.025 .. EXISTE;G CO'. UCTION TO 3E -

                                     '.-DEFLECTION OF TE RESTRAINT                                                               MODIFIED -               .0                o > . 044 IN ANY RESTRAETED DT"" cTTOL                                                                                                      w i

DUE TO THE MASS OF THE RESTRAE;T 0I

                                     .AND PIPE, MUST NOT EICEED .025".                                                                                                                                                                   J.

j

                                                                                                                                                                                                                                  . 'y

\.- . . . . 7' ' .

                                                                                                          ..                          LOAD APPLIED IN RS5T.                                                                               5
                                                                                                                  -              ,-             DIRECT""

2 AMPI.E: T " .7, ANCH. 1 " BANGER A Z:D + (10 A 5A 10)C "' 2

               - ..*pp                                                                                    .
                                                                                                                                                                                                                                      .g..

N x - BANGER B Z:E + (5 + 10 2 + 15)C "'4b ,[ 10' - ~,. s.

                                                                                                                                               *Y:E + (10-42.

15

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                                                                                                                                                                                              . .       oY                               !

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                                                                                             ~ l?                              '

• ASSUMING ANCH 1 SUPPORTS '

5' - 1007. OF RISER ONLY. s, .

                                                                                          .      C.           .
                                    'M                                                                                                                 11;,                                                                         -

1, I FIPE WT = C E

                                                                                                                      .WT. HANGER A = D lb                                                       ,'                                               l WT. HANGER B = E'lb y                                                                                                                                        . , -

j aen. 2 g .y,. m, awn 7 . u.t u .' -t tus 0F IM IO ogf te creo pee ceA/wa 1 n.q.3 . [\(\ y O'Q '

                                                                                                                                                                                            #4 ./8 RESTRAINED DIRECTION

• Q .. QA \] '

PIeE wr. + urscER wT. 1 A. g 9

                                      .. SEE NOTE 3
                                                                                                                                                          /                                  / 2.f'76                                            ,

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ATTACHMENT R

!                                                            DESIGN GUIDELINES PROVIDED TO EARTHQUAKE' ENGINEERING SYSTEMS DATED OCTOBER 27, 1977 4

i i O I a mw m,- n ~

• e -.-e,,-,--n,,-..-m r,m,-,we,-,,r-r-~, ,,,-ve-,,-,r-ewe-,wwe,- -- - ,,nv _ ,. -- ~~- - ~~ ~ ---- - -- - - - ,- ~ ~ --

l ATTACHMENT R i t O October 27, 1977 I Hancer Design % wimt j PG&F File No 40.2 Diablo canyon Earthquake Engineering Systems l 141 Battery Street, suite 400 Sa:a Francisco, CA 94111 Attention: Mr. Nat Otauhan Gentlemen: Please find attached Revision 1 of the " Guidelines for Design of l Pipe Supports and Restraints."

                                      ':he revision consists of, in brief, clarification of File 33 loads, l                        allowable stresses for special sections, and the addition of Mos?ri loads for File 44 supports attached to the Contairunent Interior Structure and the containrnent annulus steel.

l should you have any questions or corseents on the guidelines, please ! do not hesitate to contact the undersigned. Very truly yours, D. V. KELLY Chief Mechanical & Nuclear Engineer l By: D. J. Curtis DJC/lsv Attachnents i oci ALomn l RI2acher HJOormly . IFuall AGwalther O a-1 P D 02532

        =

l Guidelines for Design of Pipe Supports and Restraints Revision 1 1.0 Scope 1.1 The following general guidelines are issued to provide criteria for design of pipe supports for the Diablo Canyon Power Plant. Thic document is not meant to take the place of applicable codes and standards such as ANSI B31.1, the MSS Standards SP-58 and SP-69, AISC Manual of Steel Construction, ASME Boiler and Pressure Vesscl Code Section III and the catalogs and manuals published by component manufacturers. It is the intention of this document to guide the designer in load combinations, maximum stress allowables, and other physical requirements of pipo restraints. 1.2 Circumstances may be encountered which this document does not cover. For those subjects the Designer should refer to ANSI B31.1, AISC Manual of Steel Construction, ASME Boiler and Pressure Vessel Code Section III, the catalogs and manuals published by component manufacturers, or other appropriate publications for guidance. 2.0 Physical Requirements 2.1 Natural Frequency a) The natural frequency of a seismic restraint with its tributary pipe mass must be greater than 20 Herts in the pipes restrained ""' direction. The mass used to calculate the natural frequency shall include the weight of the restraint, restrained pipe, pipe insulation, fluid, pipe attachments, and valves. Any rational analysis may be used to calculate the natural frequency. An acceptable method would be to limit the deflection that will occur if the restrained mass multiplied by lg and applied in the restrained direction is limited to 0.025 inches. (Springs are not being considered as a support.) b) Tne natural frequency of a support in its unrestrained direction shall be considered for the purposes of computing loads and stresses. Only the weight of the hanger applied uniformly along its length needs to be considered. 2.2 Gaps - The maximum total gap allowed in the seismically restrained direction is 1/8". In non-restrained directions, the support design shall allow clearance for the most severe thermal and seismic movements of the pipe. Proper installation tolerances shall be provided where thermal movement cannot be accommodated within the specified gap minus 1/16". 2.3 Deflections - The maximum allowable deflection for restraints with gg,; D.E. loadings is 1/16" in the restrained direction. R-2 h P D 02533

2.4 Location - The design location of the supports in a straight run of the pipe may deviate by 1 10.D. of the supported pipe from the theoretical location provided it is not adjacent to equipment nozzles or heavy valves, in which case, prior approval from the ()

       '-                             Mechanical Engineering shall be required. If this relocation                                     l tolerance is utilized by the designer, he must specify on the hanger details the revised relocation tolerances for the hanger installations.

2.5 Spring Supports - shall be capable of exerting a supporting force equal to the load, as determined by weight-balance calculations, plus the weight of all hanger parts, such as clamps, and rods, that will be supported by a spring at the point of attachment to a pressure-retaining component or to an integral attachment. The design shall be such as to prevent complete release of the component load in the event of spring failure or misalignment. Any variability of a supporting spring force risulting from movement of the component shall be considered in the loadings used in the stress analysis of the component. The springs availabic travel will be checked against the thermal and seismic movements. 2.6 Hanger Rods - In no case shall hanger rods less than 3/8 in diameter be used for supporting pipe 2 in. and smaller, or less than 1/2 in, diameter for supporting pipe 2 in, and larger. Hanger rods shall be subjected to tensile loading only. Rod hanger assemblies shall be designed to allow anticipated thermal horizontal movement without subjerting the pipe to extraneous loads. The maximum swing angle due to horizontal pipe movement should be less than 4 0. If the swing angle of the rod is in excess of 4 and/or the total movement is in 0 excess of two inches, the hanger shall be offset two-thirds of the thermal movement towards the direction of movement. 2.7 Snubbers - The snubber assembly shall be offset two-thirds of the thermal movement in the cold position if the swing angle exceeds 50 and/or the total movement of the point of attachment on the pipe is excess of two inches. The midpoint of thermal travel for snubber strokes shall be set ut the midpoint of the total travel with hot and cold settings established accordingly. 3.0 Restra' int Loads 3.1 The loads applied from any source shall be considered acting in the theoretical directions plus or minus the maximum field installation tolerances unless special instructions are made on the drawing. 3.2 The loadings that shall be taken into account in designing a component support, but are not limited to, the following:

                                     ' a) Weight of the component and normal contents under operating conditions.   (DL)

' b) Superimposed loads and reactions induced by the supported system components. These include loads created by fast valve closure (TV) or relief valve (RVOT). O f R-3 L

e. . _ . . n- . - , . . - , , , . -

c) Weight of the component and contents under test conditions. (i.e. Hydrostatic Test Load-HT) d) Loads generated by restrained thermal expansion. These include g temperatures at normal operating conditions (TH) and accident conditions (THA) e) Loads generated by anchor or support movement. f) Seismic Loads - Loads due to Design Earthquake (DE), Double Design Earthquake (DnE), and Hosgri (HOS) , Double Design Earthquake loads shall be twice the Design Earthquake Load. The design seismic load for Class I pipe supporte and restraints for piping not seismically analyzed by a detailed dynamic analysis, (File 44),;i .I may be found in Appendix A. g) Friction Loads - Friction loads (FL) are to be applied in the direction of thermal movement. Its magnitude shall be the friction coefficient times the pipes dead load. The friction coefficient for steel on steel shall be at least 0.3 and for Teflon 0.07. 3.3 Loads used to prepare calculations shall be recorded on a one page summary sheet. Reference documents used to obtain loads shall be j referenced on the sunmary sheet. This sheet shall be termed " File i 33." I I 3.4 Seismic Loads in File 33 shall be represented in two categories - X-Y earthquakes and Z-Y earthquakes. The highest component among the i two shall be used to design the support. In the case of anchors, the' highest components among the two for each pipinc analisis shall be added to those from the other side of the anchvc to arrive at the resultant seismic load. 4.0 Allowable Stress for Support Structural Components 4.1 The following four load combinations will be used to compare actual stresses with allowable stresses. Load Case Loads

1. TH + DL + FL

2. TH + DL + FV + RVOT + DE

3. THA + DL + FV + RVOT + DOE

4. HOS + DL

           .where:

DE - Design Earthquake DDE - Double Design Earthquake DL - pipe Deadload FL - Friction Load FV - Fast Valvo Closure Load HOS - Hosgri Earthquake R-4 h

          ,                                                  _4_

RVOT - Relief Valve - Open. System (Transient) TH - Restrained Thermal Expansion Loads from operating conditions THA - Restrained Thermal' Expansion Leads from accident conditions . (When accident condition thermal loads are noner.istent, j TH (normal condition thermal load) shall be used in place l g o. ; of THA to evaluate load case #3.) l

                                                                                          ~

4.2 Stresses shall not exceed the appropriate maximuu values shown in Table 1. - LOAD CASE'(FROM SEC. 4.1) STRESS 1 2 3 4 a Tension 0.35 Fy 0.417 Fy 0.9 Fy 07 Shear 0.58 Fv 0.694 Fv 1.44 Fv 1.44 Fv Compression 0.58 Fa 0.694 Fa Fa or 2/3 Fcr For or 1.2 Fy 0.44 Fy 0.52 Fy 0.9 Fy Crippling Fy or 0.7 Fu

Bending 0.58 Fb 0.694 Fb 1.5 Fb 1.88 Fb Bearing 0.52 Fy 0.625 Fy

[ Tension 0.58T 0.694T 1.5 T 1.88 T N m -- Shear 0.58V 0.694 V 1.5 V 1.88 v Concrete Dwg. 054162 Dwg. 054162 2X (Dwg. 054162 2X (Dwg. 054162 Expansion Table A Table A Table A) Table A) Anchors Catalog Catalog Catalog 2X (Catalog 2X (Catalog Items values values values) values) TABLE 1 Notes 1. Nomenclature for ellowable values are from the AISC Steel Construction Manual, Seventh Edition except as follows: T - Allowable tension on a bolt per AISC V - Allowable shear on a bolt per AISC Fcr - Critical buckling strength O a-s

2. Minimum yisld ctrengths (Fy) chall b3 taken at tbn rupportzd pipss maximum operating temperature for pipe attachments. The supports

                                                                                         ", ' I temperature may assume to drop to ambient emperaturcs at a rate of I

1000F per inch from the pipe attachment.

3. Concrete Fasteners - The allowable loads and design procedures for O

embedded concrete fasteners and expansion bolts shall be in conformance with PG&C Standard 049324 and 054162.

4. Springs, Snubbers, and other Purchased Support Components - The allowable load recommended by the manufacturers of the springs, snubbers, and other purchased support components shall be used for pipe hanger designs. .

5. Stresses shall be combined as in Section 1.6 of the AISC Manual of Steel Construction substituting the allowable stresses of this table with the appropriate allowable valves from the manual.

6. Allowable Stresses in angles shall not exceed those shown in Table 1 for Load Case 2 regardl,ess of the loading combination.

f.v !

7. Allowable bending stresses for "non-compact" sections shall be limited to 0.9 Fy. "Non-Compact" sections, as defined herein, are sections i defined by AISC 1.5, 1.4.4, 1.5.1.4.5 and 1.5.1.4.6. l 4.3 Welding Requirements 4.3.1 All materials shall be of weldable quality and shall be of the same type or of compatible composition with the material to which it is to be welded and of comparable yield and tensile strength or higher.

4.3.2 Welds shall be proportioned to meet the stress requirements given in Section 4.2. For welded attachments to the pipe, li s I the allowable loads given in PFI Standard ES-26 should be used. 4.3.3 Attachments velded to PG&E Code Class A piping shall be welded with complete penetration welds contour per Figure 1-727.4.4 (a), B31.7. 4.3.4 The minimum fillet weld size, except for fillet welds used to reinforce groove welds, shall be as shown in Table 2, .' gg, g except that the weld size need not exceed the thickness of the thinner part joined. Material Thickness of Minimum Size Thicker Part Jointed of Filled Weld To 1/4" Inclusive 1/8" Over 1/4" to 1/2" 3/16" Over 1/2" to 3/4" 1/4" Over 3/4" to 1-1/2" 5/16" Over 1-1/2" to 2-1/4" 3/8" Over 2-1/4" to 6" 1/2" h Over 6" R-6 5/8" TABLE 2

                                      .e_-

4.3.5 Tha maximum fillst wald siza permitted along edges of material shall be: fs 1. The thickness of the base metal, for met'al less than (,) 1/4 in, thick.

2. 1/16 in. less than the thickness of base metal, for metal 1/4 in, or more in thickness, unless the weld-is designated on the drawing to be built out to obtain full throat thickness.

4.3.6 The minimum effective length of a strength fillet weld shall be not less than 4 times the nominal size, or else the size of the weld shall be considered not to exceed one-fourth of its effective length. 4.3.7 Intermittent fillet welds may be used to transfer calculated stress across a joint or faying surfaces when the strength required is less than that developed by a continuous fillet weld of the smallest permitted size, and to join components of built-up members. The effective length of any segment of intermittent fillet welding shall be not less than 4 times the weld-size with a minimum of 1-1/2 inches. R-7

App:ndix A File 44 - Seismic load Deternination The amount of weight to be applied at each restraint should be one-half the weight of the uniformly distributed load between two adjacent restraints acting in the same restraining direction, plus a proportional amou..t of the concentrated weight between those two restraints acting in the same restraining direction, multiplied by a seismic coefficient. The amount of concentrated weight to be used should be equal to g where A = Distance from the concentrated weight to the next adjacent restraint acting in the same direction. L = Distance between restraints acting in the same direction on cach side of the concentrated weight. W = Weight of concentrated element. Uniform loads include, but are not limited to, the weight of the pipe, pipe insulation, and fluid. Concentrated weights include, but are not limited to, valves, flanges, pipe attachments, flow elements, and half the length of branch lines betweer. the branch and the next adjacent restraint acting in the same direction on the branch line. The concentrated weights will be shown on the hanger isometrics. Followir.g is an example of weight distribution. Please refer to Sketch 1. Z Tributary Weight llangor A = D 4 [l 5 + 10 V+ 2 / N liangerB=E+f5 10

                                                     +

2 lianger C = P + # U 5 + 10/ 2 X Tributary Weight liangerA=D+f5+10 + " Ifanger B = 0

                !! anger C = F +              P+    10 + 5 + 7 +     N where D, E, and F = weight of restraint R-8

V = waight of valve P = weight per foot of branch line N = weight por foot of main pipe seismic coefficients shall be selected fro:n Table 1. For the Horgri analys,s, the annulus structure is divided into five (5) sections at elevations 106 and 116 as shown in Shotch #2. For positions where pipe supports are attached to elevationsand sections between those shown in Table 1, ' linear interpolation shall be used to' calculate the appropriate design load.  !- i i i .i ,. Sketch 2 f= t ! t

                                                                                                                    -        i.

I

! . + j

                                              .                          FRAME                    g                          l It %
'                                                                                                                            s

' l A  :

                                                                                + _ _ _. y w ..e s                _

n

                                                                             %*                                 l i                                                                                                 -.c FRAV.55 b
                                                                                                                              ^
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• P t' -9 m- r v -w ,.=.i ._r_... , , , , , , , , . _ _ . , , _ _ , _ _ , _

9 TABLE 1 Earthquake Building Elevation Seismic Coef ficier.t Horizontal l Vertical' Auxiliary All 2 1 DC Containment All 2 1 Auxiliary 100' and below 2 1 DDE - above 100' 4 1 Containment All 2 1 Hosgri Auxiliary 100' and below 2 2 140' to 115' 4 2 163' 5 2 Containnent Below 130' 2 1 (Exterior) 130' 2 2 and above 106 and below 2 1 i Containment Interior 140 3.5 1 101 1.11 4.0 106 Frame 1 1.5 2.5 Containment Frame 2 1.5 2.5 , Annulus Frame 3 1.5 16.0 Structure Frame 4 1.5 3.0 Frame 5 1.5 5.0 116 ' Frame 1 2.0 5.0 Frame 2 2.0 3.5 i Frame 3 2.0 12.0 . Frame 4 2.0 1.5 i Frame 5 2.0 4.0 140 3.5 8.0 R-10 h

Js s N

                         /Y                            -

y ,, 1 Z X j Valvo i /

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     ,                                 e llanger A r,. .%                                                                    .
                  ,L s                         # 11 Alt LI;;E
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R-11 . SKETc! 1 4

i 1 O l l l l .I ATTACHMENT S l DEFINITION OF DRAWING LABELS l l 1 0

.. 4 o c. .m. e.ec. ATTACHMENT S PGwE p'OR WTU.% .= f*CM P A NV LJ'18"n civir.sen o, PROJECT ENGINEER ocnntuznr Units 1 & 2 - Diablo Canyon Site Fitr. fio. 1.30 Q LETTER or ener " Preliminary Copy" Drawings, " Advance Copy" Drawings,

                   " Expedited Copy" Drawings, Architectural, Electrical, Mechanical and Structural Revision Sheets September 30, 1971 MR. M. H. CHANDLER:

Attention: R. S. Bain I believe that the meaning of the labels " Advance Copy" Drawing,

            " Preliminary Copy" Drawing, and Architectural, Electrical, Mechanical and Structural Revision sheets do not have a com=on meaning to all concerned.

I understand Ouality Assurance Procedure PRE-3 is being revised to include definitions for these terms. In the meantime we plan to use the following definitions (Please note we have added a new term " Expedited Copy".): PRELIMINARY COPY This is a drawing which is still under preparation. This drawing is for information only and can be used for material procurement. This drawing should have a " Preliminary" stamp above the title block. ADVANCE COPY This drawing has been coordinated, signed by the Responsible Engineer and is being reviewed by the Supervising Engineer for approval and issue. This drawing is issued for installation. Concrete cannot be poured with an " Advance Copy" drawing. This drawing should have an " Advance Copy" stamp above the title block. EXPEDITED COPY This drawing has been coordinated, signed by the Responsible Engineer and the Supervising Engineer. This drawing is " Approved for Construction". This drawing should have an " Expedited Copy" stamp above the title block. ARCHITECTURAL, ELECTRICAL MECHANICAL AND STRUCTURAI, REVISION SHEETS These sheets display a revised portion of an approved drawing (or a sketch which will be transf erred to a drawing) . These sheets are used to expedite information to General Construction on important changes thar,will be incorporated into the next drawing revision. These sheets have the same status as an

            " Expedited Copy" in that these sheets have been coordinated (if required), signed by the Responsible Engineer and the Supervising Engineer. These sheets are approved for construction and will appear on the next revision of the drawing to which it refers.
                                                                                *. . L L /{l  h p                                                                     3. f. .

O. SCHUYLER V JOS/ mas cc: FD!aut : /RVB e t t in g er /DVKelly/ KLCDor king /J RHer r era /WRH er s ey/WJLindb lad D N i els e n /D LPf l ey / RHRa ns d a ll / GVR icha rd s / EPWo lla k/JWo odwa rd S-1

O ATTACHMENT T PGSE RESPONSE TO NRC BULLETIN 79-94 DATED APRIL 17, 1980 l d O

l -

   .i                                                                                                                   ATTACHMENT T
      =ACIFIC                      GAS ANE                             E u E C ' : ' =.I C                     C C M _=ANY
    ;,A w : Epa A?wtN T . 77 3,E A l.I $ 7 3 t ET, 3 : 37 F t.C C R

• SAN
• t A N C I S C C , O A L.! F C N N I A 94:06 . (4:2) 731 4:::

April 17, 19c0 M.r. R. H. Ingelken, Directer

• Office of Inspeccion and Inferce::ent 7egica V '

U. S. Nuclear Regulatery C - issic 1990 H. Calidernia 3culevard 2 l Walnut C:sek Plass, Suite 202 Weinut Creek, Califer=is 9459d - Re: Decket No. 50-275 Decket No. 50-323 Diablo Canyen Units 1 and 2

            ~

Reference:

U. S. NRC 3ulletin 79-14 dated July 2, 1979 Des: Mr. .-.ge

• lken : ,,

O.is letter is written in respense tc the referenced II 3u11cti.n and supplements which requi c a pi - " ' specticn to verify ~ that actual ccnfie.uratiens of safetv-related .=i=ine. . . ag cc vid the

        =ed els used to sei==i=s11.v ansi.v. ce th==.

Cu- latter of Cc:cher 17, 1979 sub d' v cur respcase to te:: 1 cf the bulletin. rni.s letta:( respends ~ -s 2 , 3, and 4 fer Disblo Canycn Uni: 1 and is cur fi.21 respense fer 212b10 m :y= = Cnit 1. Review of Diabic Canyen Unit 2 and the attendant repc.. will follev at a later date. Ita=s 2 and 3 call fer a field ins,eecticn of as-built .ci=. ine. cenfic.uratices , a ec==.e. risen with the cer es.x= disc. sei==ic an21.vsis

        =cd e l s , and a descriptien cf the results of the review. Iter: 2 re-quests that this be done for nc- ' 'y accessible sv. ster:s , and Ita:: 3 requests the ase.e acticn fer inaccessible systecas. Since Disbic Canv.en is net yet an operatin= .clant all syster:s are accessible and no distinctic= w :s =.ade between It- 2 and 3. s.v s t e=s represected by a tet.21 ef 75 piping iact:etrics ecvering all seis=ic Class 1 piping I

s 2-1/2 inches and in ger and 54 inc=etrics ccvering all ae' 3-y alyced 2-inch a.d s= aller piping vere inspected by field percen cl.

        .ncy were instcuc ed t.: yellev cve. ite::s in agree =ent with the draw-l        ings and to r2rk in red all differing infc. ::icn. Tnis inscec ice i

T-1

l i

   ,      y.r . p.. E. F.ngelken                                                        2                                        April 17, 1950 1

I was ec=pleted en Nev -ker 9, 19 9. Fellcui., the inspectien, the isc=et-ics Vere Occpared

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                                                                                      . - -  - aaz=e  ,   e-.     ,4                 se4 s t. n analyses. Sece of de piping, such as the reacter coclant lecps, was analysed by Westinghcuse, which perf:=ed de 79-14 review f these sv. s te=s . F:: each a.alysis, a .rackac. e was .=re. cared Oc=. arine.

nupport 1ccaticas and directions, pir. e gec=e trv , valve data, =ateria: . prcperties and other infer =atien. A list of discrepancies was cc=pi. and each discrec.ancv was reviewed ' by a tea = cf en,ineers to assess i' impact and te de:er=ine an apprcpriate ccurse cf acti n. ""his revie v a a- c ... ' e *- ~' :' ek - . .- 1~s,

                                                                        .'sc0.

The f=11cving types cf discrepancies are typical cd dcse f:end , in crder of frequency of eccur ence : valve weights === cer-rec:t weights of valve flanges not ==d ell ed ; center c' gravity of valve operater nc ad ec.tately cens id ered ,- supper: location differene: of greater than ene pipe dia=e:er; succc--ts =is sing er extra; .=r e s e n : .. c .'. %..' ,%.. -' e..s

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c.. u:.. , "-.a - e s n.' ~.s v e - a- . . ~ ~ c wvd c u s -' ," -. . . . . c . "-.d v e .= .' * '.a . ^-. . e - cheuld be reanalysed. ~nis a=cented'to a.pr:x d - a'1" a 2 6 per~26: reanalysis rate. In additien, there were 10 large dienetor and 4

         = 211 diameter analy=es f=r which differences were roccived by a field hardware chanc.e. For the Westin,hcuse--analysed lines , althcug}

cc .a =i=cr dLss,reements hct.seen the analysed a.d the a:-built cendl-tien were f=end , Westinc.heus e deter =ined that =cne of, dese disac. rec-

         =ents were scricus ancugh := unrranz reanalysis.                                                                             .

Itec 4 requires a. c~ealu.ati=n cf tha effect of the dis-crerancies en s.ys:c=1 cr. era.bility o a achedule for e=:pletion cf re-a n a l .v o i s , and a description of precedures in effect to assure ti=cly [ incorporation cf as-built changes. S.e reanalysis was cc=pleted en i , A.=ril 11, 1980. "he results of the rea.nalvsis were as fellevs : . Oso of t':e analvses shcwed an everstress in'the .ci:.e due te de dis-1 crepancies found, a .d additicnzl suppcrts vera added tc ecrrect de pr:ble=. Tselve of the analyses chcues ne significant increase in j lead, deflecti n er stress. O.e rc=aining 43 analyses h.ad nc probler with respect te pipe stress but had i.ncreases d- per reactions. At this t i.-: , all su .=crts which have increased desi,n lead are boir

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c..ecx ed . ., -- d e ., .. 4 e wh e .s. o At present chcut 20 suppcrts have been identified as needing reve%

         ..oevaluation cf suppcrts will be cecpleted ---- - xir.ately F. y 1,                                                              1 W.

Field =cdi'icati ns vill be finished appr x'---telv~ Jena 1, 1950. l T-2 t l l

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__ 4,4A ._m_awa4 __.s i O i t 3 i ATTACHMENT U VIEWGRAPHS OF RICHARD B. HUBBARD FOR LIMITED APPEARANCE STATEMENT BEFORE [ j ASLB DATED OCTOBER 18, 1977  ! t i I I F l I

      -O

($) RISK UNCERTAINTY - DIABLO CANYON NUCLEAR UNITS QA IMPLEMENTATION DEFICIENCIES

• INADEQUATE DESIGN CONTROL SYSTEM e NO HIERARCHIAL PLANT PARTS LIST e RECENT I & E CITATIONS DESIGN CLASSIFICATION INCONSISTENCIES FAILURE TO DOCUMENT DESIGN REVIEWS FAILURE TO RECORD DESIGN CHANGES S NO P G & E SEISMIC DESIGN GUIDE
• PROCUREMENT CONTROL LACKING e WESTINGHOUSE REQUIREMENTS NOT DEFINED P.O. NOT UPDATED EQUIPMENT SHIPPED PRIOR TO SEISMIC REQUIREMENTS 9 CLASS 1 VALVE SPECIFICATION INADEQUATE "0"G VERTICAL LOADING
• RESULT - DESIGN DEFINITION LACKS DISCIPLINE u-1

(]) Ras 101877-2

O RISK UNCERTAINTY - DIABLO CANYON NUCLEAR UNITS ASSESSMENT OF QUALITY ASSURANCE PROGRAM

• BASIS OF REVIEW e NRC I & E REPORTS e FSAR 4 REGULATIONS & REGULATORY GUIDES S SER
• RESTRAINTS S NO IN-PLANT DISCOVERY e GA0 REPORT INCOMPLETE
• RISK ASSESSMENT UNCERTAINTY 9 INFLUENCED BY QA IMPLEMENTATION I

l l l l U-2 (II RBH 101877-1 1

RISK UNCERTAINTY - DIABLO CANYON NUCLEAR UNITS (3> EQUIPMENT DEFICIENCIES DISCOVERED LATE IN CONSTRUCTION

• PROCESS CONTROL FLAWS S FAULTY WELD - STEAM GENERATOR INTERRUPTED PREHEAT AT END OF SHIFT PREHEAT STOPPED BEFORE STRESS RELIEF CRACKS ON ID OF N0ZZLE DUE TO GRINDING e I & E CITATION - RADIOGRAPHS ONE RADIOGRAPH - MARCH OTHERS BY SAME REVIEWER - JUNE 3 BY OTHER REVIEWERS - JULY 13 RADIOGRAPHED WRONG WELD - JULY 13 REPAIR - 62 OR 1675 WELDS - AUGUST 4 0 NO FORMAL TRAINING PROGRAM FOR QA INSPECTORS
• INSPECTION PROGRAM DEFICIENT 4 I & E CITATION - PIPE HANGER & SUPPORTS WELDS REREVIEW ALL CLASS 1 SUPPORTS - APRIL 76 OVER ONE YEAR OF REPAIRS 9 IMPROPERLY INSTALLED ANCHOR BOLTS l REPAIR 20% OF ANCHORS - 2600
• POTENTIAL DEFECTS - CONTAINMENT ELECTRICAL PENETRATIONS e FIELD FAILURES & FIELD FIXES
• IDENTIFIED QA DEFECTS MAY BE TIP 0F ICEBERG U-3 RBH101877-3

(]) l L __ .__m

RISK UNCERTAINTY - DIABLO CANYON NUCLEAR UNITS O QA PROGRAM FOR OPERATIONS VIOLATES NRC REQUIREMENTS (CHAPTER 17.2 0F FSAR)

• FSAR DESCRIPTION INADEQUATE e DOES NOT DESCRIBE "HOW" IMPLEMENTED 10CFR50.34B REG GUIDE 1.70 e INDEPENDENT ASSESSMENT NOT POSSIBLE
• OPERATIONS QA REVIEWED TO 1972 REVIEW PLAN 0 DOES NOT MEET CURRENT STANDARDS PERSONNEL QUALIFICATIONS TRACEABILITY INSPECTION AND TEST CONTROL TEST EQUIPMENT NONCONFORMANCES CORRECTIVE ACTION QA RECORDS e REVISE FSAR OR TOPICAL REPORT
• RESULT - OPERATION QA NEEDS RE-REVIEW u-a RBH101877-4 l

O RISK UNCERTAINTY - DIABLO CANYON NUCLEAR UNITS EFFECTIVENESS OF NRC I & E PROGRAM QUESTIONED

• GA0 EXPRESSES SAFETY CONCERNS e FACT OF GAO AUDIT NOT DISCLOSED TO ACRS S GAO AUDIT - MAY 30 TO JUNE 10, 1977 NRC NOTIFIED JUNE 24 & JULY 22 NRC FOLLOWUP WITH APPLICANT - JULY 11 SOME ALLEGATIONS SUBSTANTIATED GAO REPORT ISSUE - FEBRUARY 1978
• STUDY RECOMMENDS QA IMPROVEMENTS 9 ADEQUACY OF I & E PROGRAM BROWNS FERRY NORTH ANNA DAVIS BESSE 0- RESPONSE - SANDIA STUDY INITIATED IN 1976 RESULTS RELEASED SEPTEMBER 1977 "FURTHER IMPROVEMENTS WARRANTED"
• RESULT - DIABLO DA BREAKDOWNS NOT UNIQUE O "-5 RBH101877-5

O RISK UNCERTAINTY - DIABLO CANYON NUCLEAR UNITS

SUMMARY

AND RECOMMENDATIONS

• ASLB FINDINGS NOT WARRANTED, UNTIL e INDEPENDENT ASSESSMENT OF QA IMPLEMENTATION IDENTIFIED FINDINGS MAY BE TIP OF ICEBERG REVIEW GA0 ASSESSMENT e OPERATIONS QA DESCRIPTION REVISED REVISE FSAR OR ISSUE TOPICAL REPORT NRC REVIEW TO CURRENT REQUIREMENTS e I & E EFFECTIVENESS DETERMINED l REVIEW RECOMMENDATIONS OF SANDIA STUDY e WHY GAO STUDY NOT DISCLOSED TO ACRS U-6 RBH101877-6 ll>

i 4

 ]

l, l lO i f i 4 a i i 1 ) ATTACHMENT V 1 LIMITED APPEARANCE STATEMENT OF RICHARD B. liUBBARD 4 } BEFORE DI ABLO - CANYON ASLB DATED OCTOBER 18, 1977 i i i t i i i 4 1 i i l 1 O

O i i 4 DEFICIENCIES IN DIABLO CANYON QUALITY ASSURANCE (QA) PROGRAM i AND IN QA PROGRAM IMPLEMENTATION .,i 4

TESTIMONY OF RICHARD B. HUBBARD i

1 ATOMIC SAFETY & LICENSING BOARD OCTOBER 18, 1977 AVILA BEACH, CALIFORNIA 1 i i O r i

                                                                           ,,n.-,       r----,m--e-~e.-.-   - - - - - - -

DEFICIENCIES IN DIABLO CANYON ^UALITY ASSURANCE (QA) PROGRAM g AND IN QA PROGRAM IMPLEMENTATION

SUMMARY

AND RECOMMENDATIONS The "af ter the fact" discovery by the NRC of quality de-ficiencies at the North Anna Plant, Browns Ferry Plant, and Davis Besse Plant raise serious questions about the adequacy of the whole NRC Inspection and Enforcement Program. In particular, questions need to be answered about the NRC policy of relying on builders for primary inspections with NRC officials serving as aud itors . Atomic Safety & Licensing Board findings on the Diablo Canyon non-seismic design should not be issued until the NRC Staff and the Applicant complete, as a minimum, the following four required safety assessment and modification steps as described in detail in this testimony:

1. Numerous instances where implementation of the Diablo Canyon QA Program has been demonstrated to be ineffective are described including examples of failures to control procurement documents, to properly interpret radiographs , to control special processes such as welding, and to control design records. Of equal significance, is the fact that these QA deficiencies were not discovered (by the Applicant or the NRC) until the construction of the plants was essentially complete. A thorough independent assessment of Diablo Canyon QA Program effectiveness appears warranted.

O i

O 2. A number of instances are identified where the Diablo Canyon QA Program for Operations, as described in Section 17.2 of the FSAR, is not in compliance with the requirement of the NRC Regulations as described in 10 CFR 50 and the methods des-cribed in the NRC Regulatory Guide 1.70 - 1975. As a minimum, a reassessment of the Diablo Canyon QA Program for Operations by the NRC is required.

3. The Government Accounting Office (GAO) conducted an independent assessment of the NRC inspection program at the Diablo Canyon facilities between May 30 and June 10, 1977. The GAO audit identified numerous allegations of improper workman-ship. The question sh'ould be answered by the NRC and the Appli-cant as to why the fact of the GAO audit in proct $s was not disclosed to the ACRS at the August.2 and August 12 meetings where the Diablo Canyon QA program were reviewed. The ASLB should also request GAO to review the GAO conclusions with the Board.

4. Criticisms of the effectiveness of NRC Inspection and Enforcement Program have repeatedly occurred in the past. The recommendations of the recent Sandia study (NUREG-0321) of the NRC QA Program should be obtained and reviewed by the ASLB.

The breakdowns in the Diablo Canyon QA Program and its imple-mentation identified in this testimony may be "only the tip of the iceberg." In any case, the record would appear to indicate that the NRC's and the Applicant's quality actions to date have been 11 1

inadequate to protect the public health and safety and the environment. An independent observer is left with the sense that the focus of the NRC's evaluation of the Diablo Canyon QA Program appears to have been to " prove" the plant safe enough I for operation---not an open assessment of the plant's true safety. ASLB acceptance of the Diablo Canyon QA Program at this time is not warranted. O

 .                                                      iii

4 t 1

i. .

4 9 . i TABLE OF CONTENTS i i  ! 4 I. QUALITY ASSURANCE FLAWS CAN REDUCE SAFETY MARGIN....... 1 II. QA PROGRAM IMPLEMENTATION FLAWS CONTINUE TO BE DISCOVERED AFTER CONSTRUCTION COMPLETE. . . . . . . . . . . . . . . . . 2 i A. INADEQUATE DESIGN CONTROL SYSTEM. . . . . . . . . . . . . . . . . . . . 3 ) 1. Regulatory Violation - Design Classification

Inconsistencies................................ 3 i

2. Regulatory Violation - Failure to Document De s i gn Re vi ews . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 i

3. Regulatory Violation - Failure to Record j Design Changes................................. 4  !

l B. FAILURE TO CONTROL PROCUREMENT DOCUMENTS . . . . . . . . . . . 4 j C. SPECIAL PROCESSES NOT CONTROLLED PROPERLY.......... 6

1. Quality De ficiency - Faulty Weld. . . . . . . . . . . . . . . 6

2. Regulatory Violation - Interpretation of i Radiographs............................ ........ 6

3. No Formal Training Program for QA Inspectors.... 7 D. INSPECTION PROGRAM INADEQUATE . . . . . . . . . . . . . . . . . . . . . . 7

1. Regulatory Violation - Pipe Hangers and Supports 7 '

2. Defective Concre te Expansion Anchors . . . . . . . . . . . 8 4

? E. FAILURE TO CONTROL INSPECTION DOCUMENTS............. 8

1. Regulatory Violation - Pipe Rupture l Re s t r ain t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 l F. DEFECTS IN CONTAINMENT ELECTRICAL PENETRATION SEALS. 9 i

j III. DIABLO CANYON QA PROGRAM FOR OPERATIONS DOES NOT COMPLY WITH NRC. REGULATIONS................................... 10 t ! A. FSAR DESCRIPTION FAILS TO DESCRIBE "HOW" THE QA PROGRAM WILL BE IMPLEMENTED......................... 11 {~ B. OPERATION QA PROGRAM REVIEWED TO 1972 REVIEW PLAN. . . 12 C. DEFICIENCIES IN THE QA PROGRAM DESCRIPTION. . . . . . . . . . 12 i iv

1. Criteria I - Organization...................... 13 ll)

2. Criteria VIII-Identification and Control of Materials, Parts and Components............... . 13

3. Criteria X and XI - Inspection and Test Control....................................... 14

4. Criteria XII - Control of Measuring and Test Equipment................................ 14

5. Criteria XV - Nonconforming Materials, Parts or Components........................... 15

6. Criteria XVI - Corrective Action. . . . . . . . . . . . . . 16

7. Criteria XVII - Quality Assurance Records..... 16

8. Criteria XVIII - Audits....................... 16 IV GAO AUDIT PROVIDES INDEPENDENT ASSESSMENT OF DIABLO CANYON QA PR0 GRAM..................................... 17 A. GAO AUDIT NOT DIS CLOSED TO ACRS . . . . . . . . . . . . . . . . . . . 17 B. SAFETY CONCERNS IDENTIFIED BY GAO. . . . . . . . . . . . . . . . . 18 V. NRC INSPECTION AND ENFORCEMENT PROGRAM INADEQUATE. . . . . 19 A. SANDIA STUDY RECOMMENDS IMPROVEMENTS.............. 19 ATTACHMENTS:

ATTACHMENT 1 - LIST OF REFERENCES ATTACHMENT 2 - RESUME OF RICHARD B. HUBBARD ATTACHMENT 3 - SEISMIC ANALYSIS FISHER CONTROLS COMPANY ATTACHMENT 4 - SEISMIC ANALYSIS FISHER CONTROLS COMPANY O V

i -i O DEFICIENCIES IN DIABLO CANYON QUALITY ASSURANCE (QA) PROGRAM AND IN QA PROGRAM IMPLEMENTATION l j TESTIMONY OF RICHARD B. HUBBARD FOR ATOMIC SAFETY & LICENSING BOARD 1 OCTOBER 18, 1977, AVILA BEACH, CALIFORNIA j i I I, QUALITY ASSURANCE FLAWS CAN REDUCE SAFETY MARGIN { i The purpose of this testimony is to update and expand on the t affidavits I submitted to the Atomic Safety & Licensing Board on l March 10, 1977 and supplemented on April 27, 1977. Major non-seismic issues remain unresolved which may seriously reduce the expected i plant safety margins. During my review of the Diablo Canyon Quality Assurance (QA) Program, as described in the FSAR, and the QA Program's implementation, as described in the audits by the Region V Office of Inspection and Enforcement of the NRC, I have discovered a number of instances where the Diablo Canyon Units 1 and 2 QA Program and the i

Program's implementation do not appear to be in compliance with the i

requirements of 10 CFR 50, Appendix B. l In addition, during routine QA audits in the final stages of

!              construction, the NRC has identified numerous deficiencies in the

. implementation of the Diablo Canyon QA Program which raise serious J 4 doubts about the adequacy of the plant's structures, systems, and V-1

components to function properly during normal and abnormal conditions. (l) A list of the major QA deficiencies at the Diablo Canyon Nuclear Units which has been identified since January 1976 are summarized in Section II of this testimony. Section III of this testimony summarizes examples where the QA Program for operations, as described in Section 17.2 of the FSAR, does not appear to be in compliance with current NRC regu-lations and regulatory practices. Section IV of this testimony deals with safety concerns identified by the General Accounting Office (GAO) during a recent audit they conducted at the Diablo Canyon facilities between May 30 and June 10, 1977. The fact that the GAO audit was not disclosed to the ACRS at either the August 2 or 12 ACRS meetings is also discussed in Section IV. Finally, in Section V, the results of the recently released Sandia assessment of the NRC QA Program are described. The identified breaches of QA discipline may be symptomatic of greater flaws in the QA Program which have yet to be identified. Since a key factor in assessing the potential risks of a nuclear plant is the assumption of a disciplined, thorough QA Program, the inadequacies in the Diablo Canyon QA Program and its implementation pose a possible significant hazard to the public health and safety and to the environment. My qualifications to reriew the Diablo Canyon QA Program are summarized in Attachment 2 of tnis testimony. II. QA PROGRAM IMPLEMENTATION FLAWS CONTINUE TO BE DISCOVERED AFTER CONSTRUCTION COMPLETE Numerous major deficiencies in the QA Program's implementation have been identified after a majority of the construction has been O V-2

 /~                               completed, which raise serious questions about the ability of the V)

Diablo Canyon structures, systems, and components to function prop-erly during and following the design based accidents. An independent reviewer is left with the sense that the Applicant and NRC inspections have identified the symptoms of quality flaws, but that the NRC has failed to address and require correction of the systematic failures in the Applicant's QA Program. Examples of major identified defi-ciencies in the Diablo Canyon QA Program implementation, discovered since January 1976 include the following: A. INADEQUATE DESIGN CONTROL SYSTEM During the discovery process of the seismic contention, I determined that the Applicant had no hierarchial drawing control system i which began with a plant master parts list or plant equipment piece list. The lack of such a " Christmas tree" type drawir; control system, and revision control system can result in a loss of interface control between engineers responsible for different aspects of the design. Appendix B. Criteria III, of 10 CFR 50 requires that " measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations."

1. Regulatory Violation - Design Clacsification Inconsistencies Two examples of the resulting design classification incon-sistencies have been discovered by the NRC in the past year. In October 1976, the NRC cited the Applicant for a violation since the wiring con-necting the steam generator level transmitters to the indicators on the hot shutdown panel were not installed in accordance with the requirements g- of the assigned safety classification.(1) Three months later, in V) V-3

January 1977, the Applicant was again cited when another electrical circuit was discovered as being installed as non-Class 1 in accordance with wire lists, but which should have been Class 1 as called out on the circuit drawing.(2) The status of the Applicant's review plan was reviewed by the NRC on June 30, 1977.(3) The two items have yet to be resolved.

2. Regulatory Violation - Failure to Document Design Reviews Contrary to 10 CFR 50, Appendix B, Criterion XVII, the Applicant's engineers performed several reviews of the Westinghouse NSSS design to verify incorporation of seismic parameters, but did noc document the scope or results of these reviews in an adequate and retrievable fashion. Criterion XVII requires that records of activities affecting quality must be identifiable and retrievable.(4)

3. Regulatory Violation - Failure to Record Desian Changes Contrary to 10 CFR 50, Appendix B, Criterion XVII, Specification No. 8827 (Furnish and Install Auxiliary Building Class I Ventilation System) in the Applicant's Engineering Department Central Files did not contain the latest change (Change 57, dated October 4, 1976) as listed in the Specification Change Notice Log. The Applicant's procedure PRE-4 requires that a complete file of specifications, with all approved changes, be located in the Engineering Department Central l Files.(5)

B. FAILURE TO CONTROL PROCUREMENT DOCUMENTS The seismic design requirements were not adequately described by the Applicant to the reactor equipment manufacturer, Westinghouse, in a timely manner. Appendix B of 10 CFR 50 requires that " design bases, and other requirements which are necessary to assure adequate V-4 g

i i 4  ! ( )- quality are suitably included or. referenced in the documents for p rocurement ." The Applicant in his nuclear steam supply system bid specification 8700 dated June 15, 1966 in Section 4.5 stated that "the equipment to be furnished will be required to meet seismic structural design criteria that have not yet been finally determined." No additional clarification of the seismic design criteria were in-l cluded by the Applicant in the purchase order to Westinghouse. Ground response spectra were sent to Westinghouse in 1967 and 1968, yet on November 25, 1968, Westinghouse wrote to the Applicant and asked for 4

        "the specific floor response spectra in order to complete final seismic design of Class 1 equipment."     The requested information was sent to Westinghouse seven months later, i.e., on June 25, 1969.      Major equip-i l        ment had already been shipped to the Diablo site., yet Westinghouse and j         the Applicant were still defining the appropriate seismic design criteria.

i In additicn, the engineers of the Applicant had no seismic specification which listed the appropriate floor response spectra for i

,        each equipment area. The unique floor response spectra were needed by

the Applicant's engineers as inputs to their purchase specifications for safety-related equipment. The result was inadequate purchase specifi-cations. For instance, in response to the Applicant's specif. cation l

No. 8729 for Design Class 1 valves, the valve supplier provides a 22-page seismic analysis based on a "0" g requirement in the vertical j plane (see Attachments 3 and 4) . l l l

C. SPECIAL PROCESSES NOT CONTROLLED PROPERLY

1. Quality Deficiency - Faulty Weld O

In March, 1977, the Applicant reported discovery of a crack in the weld connecting a feedwater pipe to the steam generator requiring the Applicant to examine and test other welds to verify the integrity of the full steam generator piping system.(6) Subsequently, the Applicant concluded that the following factors contributed to the failed weld: Interrupted preheat at end of weld shifts, preheat stopped before stress relief (31 calendar days between completion of the weld and the start of the stress relieving)(7) and small cracks on the ID of the nozzle due to grinding.(0) The item remains unresolved. On September 7, 1977, the NRC reported discovery of an indication approximately 11 inches long and over 0.1 inches deep when sectioning portions of the root of steam generator weld 1-4. Prelimin-ary indication was a lack of fusion.(9)

2. Regulatory Violation - Interpretction of Radiographs Contrary to 10 CFR 50, Appendix B, Criterion XVI and PG&E's PSAR, Paragraph XVI, the Applicant failed to record and disposition the crack indication on the reader sheet of the radiograph of steam generator and feedwater nozzle field weld 197.(10)

On June 3, 1977, the Applicant informed the NRC that "an audit was made of all radiographs accepted by the individual who approved the radiograph of field weld No. 197. Several similar discrepancies were found and the documents have been amended to record the significant indications. Repair of the welds for which the radiographs showed re-jectable indications is expected to be completed by July 1."(11) V-6 O 1

C) (Emphasis added) in audit of previously reviewed redtesteP hs for Class 1 piping welds was initiated by the Applicant. On July 13, that Applicant informed the NRC that "the audit has disclosed some cases in which radiographs may not have been interpreted correctly. The investigation has also disclosed that the weld attaching the main steam line to steam generator 1-4 was not radiographed. M adjacent factory weld was field radiographed by mistake...(12) (Emphasis added) On August 4, the Applicant informed the NRC that 62 field welds, out of 1,675 welds reviewed, will be repaired on Diablo Canyon Unit 1. In addition, 10 welds will be reradiographed to provide better quality films for detailed examination. (13) The item remains open.

3. No Formal Training Program for QA Inspectors Thorough training of inspectors would increase the ability and consistency of the inspector's actions. The NRC noted in 1976 that a formal program for training of new auditors and inspectors had not been developed.(14)

D. INSPECTION PROGRAM INADEQUATE

1. Regulatory Violation - Pipe Hangers and Supports In April 1976, the NRC cited the Applicant because most of the pipe hangers and supports inspected by the NRC contained a weld or welds which did not conform to the requireraents of Kellogg specifi-cations. Some supports did not meet requirements for spacing or shims or such requirements were not specified. The NRC inspector could find no evidence that the deviations from specifications had been identified and recorded for disposition by the responsible organizations.(15) ne V-7

{J

Applicant instituted a program to reinspect 200 representative hangers and then expanded the reinspection to include all heavily loaded suppoi s for Code 1 piping and the engineered safeguards piping.(10) (17) The item remains open.

2. Defective Concrete Expansion Anchors During a November 1976 audit by the NRC, the Applicant notified the NP.C that expansion anchors to attach pipe hangers and supports to concrete support structures were not properly installed.

Non-conforming conditions included (a) bolts that had been shortened by cutting off the bottom end, (b) expansion sleeves that had been shortened, and (c) anchor size smaller than specified. Nineteen of forty-five hanger installations inspected included non-conforming anchor bolts.(10) On Diablo Unit 1, of the 13,000 anchors re-inspected, approximately 20% of those anchors required rework to meet the acceptance criteria.(19) E. FAILURE TO CONTROL INSPECTION DOCUMENTS 10 CFR 50, Appendix B, Criterion V requires that activities affecting quality shall be prescribed by documented instructions, pro-cedures, or drawings of a type appropriate to the circumstances and that the activities shall be accomplished in accordance with these instructions, procedures, or drawings.

1. Regulatory Violation - Pipe Rupture Restraints M.W. Kellogg Engineering Specification No. ESD-243 for Pipe Rupture Restraints requires that inspections be performed to verify fitup, preheat temperature, root pass, weld completion and final visual examination and further requires that all operations be documented on a pipe restraint process sheet.

V-8

(]) Contrary to the above requirements, in August 1977 the NRC determined that the final visual inspections of field weld numbers FW-22A, 22B, 23A, 23B, 23C, 24A, 24B, 24C, 29A, 29B, 30A, 30B, 31 and 32, of restraint number Bent 9B (Drawing No. 1000111), had not been performed, as required, as evidenced by the entry "NA" in the final visual block of the associated restraint process sheets.(20) The item remains open. F. DEFECTS IN CONTAINMENT ELECTRICAL PENETRATION SEALS Containment failure in a melting accident develops direct pathways for radioactivity to escape to the environment. In describing sources of containment leakage, most discussion centers on pipes which penetrate the containment. Another important source of leakage is from failure of containment electrical penetration seals. Such failures are potentially very important. Diablo Canyon nuclear reactors each have about 45 electrical penetrations; typically 12" diameter pipes carrying in total 4000-6000 wires through the containment wall. (21) These wires power a wide range of devices from large pumps requiring high voltage and large currents to the most sensitive radiation moni-toring signals. Many of these cables are critical to safety in accident conditions. The double seals in such penetration are filled with an epoxy cement which serves as an insulator for the electrical wires and a seal to withstand the pressure differential.(22)

. This is where problems arise. Epoxy has been observed to i i

revert under conditions of sustained humidity and temperature. (Rever-sion is the process of the epoxy returning to its original fluid constituents and thus failing to provide either a seal or insulation.) l V-9 w-.. y-y,, ,,r - --r, ~~- ,. . - - - , . - , - - ,

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If this happened during an accident, the shorting of the critical g control and operating cables both would place the reactor operator in a semi-blind and helpless state and would create a leakage path for the steam, gases, and radioactivity released from the reactor Core. Since the vulnerable portion of the penetration seals represents a potential leakage area of over 100 square inches, only c fraction of the epoxy seals need revert and fail in order to create potentially large leakage rates and radiation release to the environment. There have been failures and field repairs of penetration seals at Diablo Canyon during the construction period. Accordingly, one can logically assume a relatively high probability of failure of the seals during their service life. Particularly as they age and are exposed to persistent moisture as is the case in the normal reactor environment. Taking into account the likelihood of metal / water reaction, steam explosions, and other transients involving release of steam and high temperatures, the chances appear significant that weak or aged penetration seals will experience a fatal environment which could result in a breach of the containment. Failures of containment electrical penetrations represents a significant source of common-mode failure which could have serious consequences. A thorough " review" of the design verification program for the containment electrical penetrations appears justified. III. DIABLO CANYON QA PROGRAM FOR OPERATIONS DOES NOT COMPLY WITH NRC REGULATIONS The Applicant and the Regulatory Staff have not demonstrated that the Quality Assurance (QA) Program for operations at Diablo Canyon lll V-10

(} Units 1 and 2, as described by the Applicant in Section 17.2 of the FSAR, comply with the requirements of 10 CFR 50, Appendix B, entitled,

     " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocess-ing Plants."  Nor has the Applicant demonstrated that the Diablo Canyon QA Program description in Section 17.2 of the FSAR is in accordance with the methods described in Ragulatory Guide 1.70, Chapter 17, dated October 1975, entitled, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants - LWR Edition."

A. FSAR DESCRIPTION FAILS TO DESCRIBE "HOW" THE QA PROGRAM WILL BE IMPLEMENTED The Applicant's QA Program description in Section 17.2 of the FSAR lacks specificity and, in general, fails to describe how the QA Program elements will be accomplished. The Applicant's lack of specificity violates Paragraph 611 of 10 CFR 50.34b which requires that the Applicant "shall include a description of how the applicable requirements of Appendix B will be satisfied." (Emphasis added) In addition, Regulatory Guide 1. 70, Chapter 17, requires that "the SAR address, at a minimum, each of the criteria in Appendix B in sufficient detail to enable the reviewer to determine whether and how all the re-quirements of the Appendix will be satisfied." (Emphasis added) In addition, the Applicant is in violation of Regulatory Guide 1.70, Chapter 17.2 which states that "the FSAR should describe the extent to which the operations phase QA Program will follow the guidance in WASH-1284,....in WASH-1283,....and in WASH-1309." In many SAR's being I docketed at this time, the SAR description of the QA Program is supplemented by a more detailed topical report which is written in sufficient detail to enable an independent reviewer to assess how the (Jg x QA Program is being implemented. Clearly, if a topical report is not V-11 l t

submitted by the Applicant, the FSAR description of the operations (l) QA Program must be expanded to meet both the letter and the intent of the regulations. B. OPERATION QA PROGRAM REVIEWED TO 1972 REVIEW PI.AN The NRC review of the QA Program for operations, as described by the Applicant in Section 17.2 of the Diablo Canyon FSAR, was con-ducted in accordance with the 1972 Standard Review Plan. In 1975. a much more thorough Standard Review Plan was issued by the NRC. At the August 12, 1977 ACRS meeting, the NRC acknowledged 12 specific standard review plan (1975 plan) guidelines where the Applicant's committment is not in full agreement with the current standard review plan. However, the NRC condluded that "none of the differences between it [theDiabloCanyonQAProgram and the current standard review plan are considered sufficiently significant, from a safety standpoint to warrant retrofitting."(23) The result is NRC acceptance of a QA Program that does not meet all specific guidelines of the current standard. Further examples of deficiencies in the Applicant's QA Program description are described in Section III-C of this testimony. The QA Program for operations should be "re-reviewed" by the NRC for compliance with the 1975 Standard Review Plan and other current NRC practices. A re-review to current regulations and practices is very appropriate as the QA Program being reviewed will not become in effect until core loading, at the earliest in 1978. C. DEFICIENCIES IN THE QA PROGRAM DESCRIPTION The proposed QA Program of the Applicant for plant operation, as described in Section 17.2 of the FSAR,.ts revised in Amendment 46 of the FSAR in December 1976, is not in compliance with the requiremen lll V-12 i

), (_) of 10 CFR 50, Appendix B, and the methods described in Regulatory Guide 1.70-1975. Representative examples of deficiencies in the Applicant's QA Program for plant operation include, but are not limited to, the following items which could significantly affect the public health and safety by increasing the probability of a massive release of radioactive materials.

1. Criteria I - Organization The authority of the Director, Quality Assurance, appears limited. The FSAR states that "the Director, Quality Assurance, has the responsibility and authority to stop the work, except in instances where stopping the work would involve changing power level or separating a generating unit from the Company system."

The restriction of QA authority appears to violate Appendix B of 10 CFR 50 which requires that "such persons and organizations per-forming quality assurance functions shall report to a management level such that this required authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations, are provided." Paragraph 17.1.1.2 of Regulatory Guide 1.70 requires that "the qualification requirements of the principal QA and quality control positions should be described." Section 17.2 of the FSAR is deficient in that it includes only the qualification requirements for the Director, Quality Assurance.

2. Criteria VIII- Identification and Control of Materials, Parts and Components The FSAR does not state that "the method and location

([) of the identification does not effect the function or q vlity of the item being identified" as stated in Regulatory Guide 1.70. V-13

3. Criteria X and XI - Inspection and Test Control ggg The description of the Diablo Canyon QA Program in the FS AR appears contradictory. For instance, the FSAR states that

" required inspections will not be bypassed." However, two sentences later the FSAR states the " decisions to bypass required inspections will be made with the approval of the Plant Superintendent and the Director, Quality Assurance." Similarly, the FSAR states that " tests will be con-ducted in accordance with approved written procedures," but then addes that " departures from approved test procedures will be per-mitted only under controlled conditions."

4. Criteria XII - Control of Measuring and Test Equipment The control of measuring and test equipment, as des-cribed in the FSAR, is not in accordance with the methods described in Regulatory Guide 1.70. The Applicant does not describe the

" standards" utilized for calibration. In contrast, Regulatory Guide 1.70 states that the test equipment should be " calibrated against certified equipment or reference or transfer standards having known valid relationships to nationally recognized standards." In addition, the Applicant does not describe the disciplined system "that will apply if measuring and test equipment is found out of calibration (1) for evaluating the validity of previous inspections or test results and the acceptability of items inspected or tested since the last calibration check and (2) for repeating original inspections or tests using calibrated equipment where necessary to establish acceptability of suspect items" as stated in Regulatory Guide 1.70. 3 V-14 W

l t '(]) 5. Criteria XV - Nonconformingj$aterials, Parts , or Components The Applicant does not meet the requirements for re-porting and documenting nonconformances. Regulatory Guide 1.70 states that the FSAR "should describe measures established by the Applicant (1) for contractors to report to him those nonconformances concerning departures from procurement requirements that are dispositioned 'use as is' or ' repair' and (2) to make such nonconformance reports part of the documentation required at the nuclear plant site or to include a description of the nonconformance and its disposition on certificates of conformance that are provided to the site." The Applicant fails to describe in the FSAR the system at Diablo Canyon for meeting either of the two preceding measures. The disposition of nonconformance also appears to violate Appendix B of 10 CFR 50. Disposition of " major nonconformance according to the FSAR will be referred to the Plant Staff Review Committee (PSRC) for disposition." The FSAR states on page 17.2-3 that "the PSRC will be made up of the Plant Superintendent and his key supervisory personnel. The on-site quality assurance engineer will attend and will participate in PSRC meetings. He will be consulted on matters relating to quality but will not take part in final Committee decisions. Apparently neither the quality engineer or the responsible design engineer will participate in the disposition of nonconformance. V-15

  )

6. Criteria XVI - Corrective Action A key ingredient of a corrective action QA program is the identification and analysis of conditions adverse to quality both at the construction site and at the manufacturer of safety-related equipment. NRC Regulation 10 CFR 50.55e requires that the Applicant "shall notify the Commission of each deficiency found in design and construction, which, were it to have remained uncorrected, could have affected adversely the safety of operations of the nuclear plant at any time throughout the expected lifetime of the plant." The des-cription of the Diablo Canyon QA Program in the FSAR does not acknowledge that the Applicant will meet this regulation nor does the FSAR describe the Applicant's method for passing this requirement on to its suppliers of safety-related equipment.

7. Criteria XVII - Quality Assurance Records The Applicant has not described the system to safeguard QA records. Paragraph 17.1.17.3 of Regulatory Guide 1.70 states that the FSAR "should describe the measures that establish requirements....

for protecting records from destruction by fire, flooding, tornados, j insects, and rodents and from deterioration by extremes in temperature and humidity." The Applicant should revise the FSAR to describe an acceptable system for safeguarding the vital QA records.

8. Criteria XVIII - Audits The QA audit program described in the FSAR is in violation of 10 CFR 50, Appendix B, and Regulatory Guide 1.70. The FSAR states that "the auditor is not always completely independent of the work he is auditing" which directly conflicts with the Appendix B requirement V-16

 /~'T that " audits shall be performed....by personnel not having direct V

responsibilities in the areas being audited." The FSAR also does not describe the use of audit checklists or the scheduling of audits, as stated in Chapter 17.1.18 of Regulatory Guide 1.70. IV. GAO AUDIT PROVIDES INDEPENDENT ASSESSMENT OF DIABLO CANYON QA PROGRAM The General Accounting Office (GAO) conducted an independent assessment of the effectiveness of the NRC inspection program in assuring that the Applicant's QA program committments were implemented in a thorough, desciplined manner. The safety concerns identified by GAO raise further doubts about the adequacy of the Diablo Canyon structures, systems, and components. A. GAO AUDIT NOT DISCLOSED TO ACRS The GAO conducted an audit of the NRC inspection program at the Diablo Canyon facilities during the weeks of May 30 and June 6, 1977 and conducted interviews with numerous craftsmen engaged in safety-related work. Pursuant to this audit, the GAO notified Region V of the NRC of the audit items of concern on June 24 by telephone and by letter on July 22, 1977. These items of concern identified allegations of improper workmanship at the Diablo Canyon facilities. A Region V inspector visited the site during the week of July 11 to begin an NRC investigation of the GAO findings. The NRC investigation was completed on August 3, and documented in a letter to the Applicant on August 23. The NRC inspection report and the Applicant's September 21 response to the " Notice of Violation" were finally placed in the Walnut Creek Public Document Room and transmitted to the Diablo Canyon service list () on a proximately September 30. (24) V-17

During the time period described in the preceding para- ggg graph, the ACRS Subcommittee on Diablo Canyon met in Des Plaines , Illinois on August 2 and the full ACRS met in Washington, D.C. on August 12, 1977. At both August meetings, the Diablo Canyon QA program and its implementation were agenda items. At no time during these deliberations was the ACRS, or the Intervenor's QA consultant, informed of the fact of the GAO audit in process. The ASLB should determine why the Applicant and the NRC did not inform the ACRS of the existence of the GAO audit, particularly since the ACRS repeatedly l l questioned whether the Applicant or the NRC planned to conduct any independent assessment of the Diablo Canyon QA compliance. B. SAFETY CONCERNS IDENTIFIED BY GAO The GA0 on-site audit identified safety concerns regarding (1) pipeway welding quality and structural stec - installation; (2) seismic and pipe whip restraint installation; (3) adequacy of concrete anchor bolt testing; and (4) resin filter trap welding quality. During the NRC investigation some of the allegations were substantiated, l though the majority of the allegations were shown to have been pre-viously addressed in a satisfactory manner by the Applicant or the NRC. l However, the NRC did cite the Applicant for a regulatory violation for l failure to control pipe rupture restraint documents as described in Section II-E-1 of this testimony.(25) The GAO audit report summarizing the GAO assessments of the NRC inspection program is scheduled to be released in February, 1978.(26) l l The ASLB should evaluate the GAO assessment as part of the Board's evaluation of the effectiveness of the Diablo Canyon QA Program. V-18 g l l

1 l l I Q V. NRC INSPECTION AND ENFORCEMENT PROGRAM INADEQUATE The "af ter the fact" discovery by the'NRC of quality deficiencies at the North Anna Plant,(27) Browns Ferry Plant,(28) and Davis Besse Plant raise serious questions about the adequacy of the whole NRC Inspection and Enforcement Program. In particular, questions need to be answered about the NRC policy of relying on builders for primary inspections with NRC officials serving as only auditors. A. SANDIA STUDY RECOMMENDS IMPROVEMENTS In partial response to the numerous criticisms'of.the NRC-I & E practices, in May, 1976 the NRC provided Sandia Laboratories of Albuquerque, New Mexico, with over a quarter of a million dollars of funding to conduct a comprehensive, independent assessment of'the NRC activities related to the review, approval, and inspection of quality assurance programs at commercial nuclear power plants. Specifically, the study assessed the effectiveness of the overall philosophy of the NRC QA program and the relative strengths and weaknesses of the practices employed to assure a high standard of quality assurance for nuclear reactors.(29) i The Sandia Study's final report was released as a NUREG series document in September, 1977.(30) While the 16 recommendations of the study group were carefully worded in a positive manner so as to not imply that the existing NRC I & E Program is inadequate, the message is still clear. The report states in t!.e summary that " based on the results of i our survey and the stringent demands for reactor safety, we conclude that j furcher improvements are warranted in both industry quality assurance l programs and NRC regulations of these programs." Specifically, the V-l9 i ([)

I study concludes that " routine direct NRC inspection and testing of g, hardware be increased, and that data pertinent to quality decisions made in the construction and operation of a plant be evaluated by the NRC on a routine basis. This includes the evaluation, for ex-ample, of radiographic and ultrasonic test data." In assessing the effectiveness of the NRC I & E Program, the ASLB should be provided by the NRC with testimony from the Sandia study team. The breakdowns in t.he Diablo Canyon QA Program described in this testimony indicate that the NRC's and the Applicant's quality actions to date have been inadequate to protect the public health and safety and the environment. The primary focus of the NRC QA evaluation of the Diablo Canyon structures, systems, and components appears to have been to " prove" the plant safe enough for operation---not an open assessment of the plant's true safety. V-20 g

p ATTACHMENT 1 V LIST OF REFERENCES

1. NRC Inspection Report 50-3?.? ,6-05, Page 8.

2. NRC Inspection Report 50,323/77-01, Page 5.

3. NRC Inspection Report 50-323/77-06, Page 3.

4. NRC Inspe: tion Report 50-275/77-12, Page 1.

5. NRC Inspection Report 50-275/77-12, Page 2.

6. PG&E Press Release entitled, "PG6E Reports Pipe Cracks In Test at Nuclear Plant," March 23, 1977.

7. Letter, Philip Crane to R.H. Engelken, June 3, 1977.

8. NRC Inspection Report 50-275/77-14, Page 1.

9. Telephone call, J.L. Crews to R. Hubbard, Sep. 9, 1977.

10. NRC Inspection Report 50-275/77-06, Page 1.

11. Letter, Philip Crane to R.H. Engelken, June 3, 1977.

12. Letter, Philip Crane to R.H. Engelken, July 13, 1977.

13. Letter, Philip Crane to R.H. Engelken, August 4, 1977.

14. NRC Inspection Report 50/323/76-01, Page 9.

15. NRC Inspection Report 50-275/76-06, Page 4.

16. NRC Inspection Report 50-323/76-05, Page 5.

17. NRC Inspection Report 50-275/76-14, Page 3.

18. NRC Inspection Report 50-323/76-05, Page 6.

19. NRC Inspection Report 50-323/77-03,.Page 7.

20. NRC Inspection Report 50-275/77-17, Page 9.

27. Diablo Canyon FSAR, Page 3.8-21.

22. Diablo Canyon FSAR, Figures 3.8-8, 3.8-9, and 3.8-10.

, 23. Prepared testimony of Al Garland, ACRS Meeting, Augus t 12, (]) 1977, Washington, DC. Al-1

24. NRC Inspection Report 50-275/77-17, Page 2.

25. NRC Inspection Report 50-275/77-17, Pages 1 to 13.

26. Telephone call, R,B. Hubbard to Willis Levie, October 7, 1977.

27. " Allegations of Poor Construction Practices on the North Anna Nuclear Powerplants," GAO Report EMD-77-30, June 2, 1977.

28. " Browns Ferry Nuclear Plant Fire," Hearings before the Joint Committee on Atomic Energy, dated September 16, 1975.

29. NRC Press Release No. 76-122, entitled " Independent Assessment of NRC Quality Assurance Activities Planned,"

dated May 25, 1976.

30. "A Study of the Nuclear Regulatory Commission Quality Assurance Program," NUREG-0321, published August, 1977.

Al-2 O

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SUMMARY

AND CONCLUSIONS ATTACHMENT 4 .

           -             Based on the calculatiens chown in this report, we have denonstrated that the pr$ nary steady state stresses, when combined with the inertial loading re-culting from the recponse to a ground acceleration of                                                                                                                                                         0              g. acting in the
              ~

vertical and 3 0 g. acting in the ho,rizontal pinnes simultaneously, produce combined stresecs which are safely within the yield stresses of the construe."

                                                            ~

tion materials, both in tension and in shcar. .

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Also, esiculations veri.fy that the extended parts .of each valve asser.bly have a natural frequency of vibration greater than 20 cycles per scecc.d. ,

 ~
                  .        In su::.ary, this seisnic analysis proves mathematica11 that the equipnent                                                                                                                                                                                                                               .
                  .         cupplied by F. sher Controls Ce::pany ,is capabic ,of perfoming all functions                                                                                                                                                                                                                                             ,,

intended within Pacific Gas & Eicctric Co. cpecifications.

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O ATTACHMENT W QA PROGRAM REVIEW REPORT R. F. REEDY AUDIT OF PGSE O

1 w- . QUALITY ASSURANCE REVIEW AND AUDIT REPORT PHASE I By: R. F. REEDY, INC. 2 On: SAFETY-RELATED ACTIVITIES r PERFORMED BY PACIFIC GAS AND ELECTRIC PRIOR TO JUNE, 1978 , } s k

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                                                                                                                                       ?G&E - March 3, 198I 1/99
     ._ . ._._-._.-. _._, _ ___ _ .., ,_.              _ _ . _ . , . . _ - _ _ _ . _           . _ . _ . _ . . _ . _ . . . . . - _ .                      1.__.._____._______

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                                                                                                      , ,   -g QUALITY ASSURANCE PRCGRAM REVIE7 REPORT PRASE I SAFETY RELATED ACTIVITIES PERFORMED BY PACIFIC GAS AND ELECTRIC PRIOR TO JUSI 1, 1978 Introduction     ;

Scoce: - On February 23, 1982 R. F. Reedy, Inc. ce=pleted the Quality Assurance Review and Audit of Pacific Gas and Electric (PG&E) safety related activities concerning- the Diablo Canyon Nuclear Project. . The purpose of this review and audit was to assess the . adequacy of PGaE Quality Assurance Progras prior to June, 1978 with particular e=phasis' on activities that could affect seis=ic related design. The baseline for this review and audi were the require =ents of 10CFR50, Appendix 3. PG&E Activities: PG&E' bad the responsibilities of Architect-Engineer and Construction Manager for the Diablo Canyon Project. PG&E was supported in their design activities by contracted design consultants. . . Evaluation Criteria: This Quality Assurance Review and Audit of PGaE addressed the requirements of 10CFR50, Appendi.: 3 with selected parts of ANSI N45.2.ll being considered for guidance. Follow-up ite=s were introduced that evolved frc= earlier R. F. Reedy, Inc. audits of PGAE design censultants. Methed of Review and Audit: The review and audi: was conducted in three steps: 1. In:redue: cry meetings; 2. Cuality Assurance Manual and Precedure review, and, 3. Auda: cf progra: i=ple=entatica. Steps 2 and 3 were perfer=ed to detailed edecklists, Inc 1 general questionaire was used for Step 1. The introdue:cr7 =eetings were held a: PG1E cc Dec. 17 and 18, 1981. The discipline groups visited were Quality 5 pG;? . u s s= 3, 133; 2/29

(:)- . .

                                                                                                                                   ~~

Assurance, Design-Drafting, Mechanical, Civil, Electrical, and Engineering Services. The Chief Engineers and other perecus from these groups were met to discuss review and audit approach, PG&E progra= and project status prior to

                .       June,    1978.          Attendees at these meetings are listed in Appendix A.

The Quality Assurance Manual and Precedures review was completed on Jan. 30, 1982 at R. F. Reedy, Inc. offices. . . Revisions of the Manual between 1970 and 1978 were reviewed along with applicable implementing procedures. The completed checklist from this review is included as Appendix B to this report. . The implementation audit was performed Feb. 2-23, 1982'at PGaE. This step of the review and audit was broken into three parts: Part A: General Re'uirements q and Management Control of . Quality Assurance . Part B: Design Control Part C: Follow-up Questions frcm Supplier Audits. Procedures and documentation were examined to ascertain . progra= coverage. Where procedural coverage w 'as not in , place, the design documentation was used for determining if positive though informal controls were practiced. Documentation examined was frem design activities performed prior to June 1, 1978. Later activities are to be separately covered in the Phase II review. Ccmpleted l I checklists frem Parts A, B, and C are included in Appendiz C to this report. l Cenclusions:

1. The PGaE Quality Assurance progra= for design work was not adecuate in areas of policy, precedures and implementation. The Quality Assurance organanation had insufficient prcgra= respcasibilit7 I

i 2. A general weakness existed in internal and esternal interface and docu=en controls. This questions whether appropriate design informatica was being exchanged and utilined by design groups and censultants. t

                                                                    "-2 O                                                                                            PG&E    . March 3,          1982 3/39

__ e _. ._:.___.._._._ _. . . . _ . . _ _ .. h concern if the latest Hosgri seis=ic data was - One is inputted for design analysis.

3. The design verification progra= vas not for=alised and was inconsistently i=plemented and docu=ented. This ,

included =ajor gaps in design overviews of the design approach for =echanical and other equi;=ent. Findings Proera=catic Deficiencies: ..

1. Quality Assurance as defined in the ~QA Manual was essentially an audit role. The Quality Assurance group was not assigned a prt=ary role in deter =ining QA require =ents.

2. PGLE had no procedure far assuring the cc=pleteness of .

the QA progrs= to address the require =ents of 10CFR50, Appendix B.

3. There were no provisions for docu=ent control of correspondance and design docu=ents.

4. During Phase I, there were no centrolled precedures for ,

design control, design interfaces and design r,esponsi . ' bilities. PRE-9 and PRE-10 on these subjects were released in 1979 and are to be audited during Phase II.

5. PG&E did not require design consultants to i=plement Quality Assurance require =ents.

6. Corrective action provisicas were not addressed excep:

with respect to audit deficiencies and deficiencies at the site.

7. Indcctrination and training were not addressed in the QA Manual or procedures.

S. The QA Manual contained no provisions fer PGLI

                          =anage=ent     review      of   the         QA pregrs= for status and adequacy.

I=clecentation Deficiencies:

1. PGLE =anagemen dic nc: review and assess the effectiveness of the Quali 7 Assurance Pregp1=.

                                                                                ?G&I - March 3,     1982 4/39

V(~T e No docu=entation was available to verify that PGaE reviewed the QA progra= to show that all requirenents of 10 cyp.50, Appendix B were addressed and =et. .. e Manage =ent Review co==ittees only reviewed plant

            -                operational considerations and experiernces frc= the Hu= bolt Bay Plant.            They did not review the QA
          .                  progra= for      design   and     construction'of the Diablo Canyon Plant.

e A program review by Energy, Inc. for an ASME N-sta=p (Dece=ber 22, 1975) lists =any of the same - findings that were found during this Phase I Review.

2. The PG1E audit syste= and corrective action syste= vere not effective. .

e Audit reporting and.' follow-up was not timely. Reports were issued sc=etimes three or four =onths after the audit. . e Corrective actions for audit findings were ineffective in that the same findings were found during later audits. e Corrective action verification was by re-audit only. .

                                           ~

e For=al co'rrective actions were not invoked on the -

                          .. engineering groups.

3. Design consultants were not required to implement Quality Assurance Progra=s.

e e Blu=e had first contract to require Quality Assurance in late 1977. e Responsible engineers did not document Quality Assurance require =ents for purebase specifications en censultants, as required by the Qualit7 Assurance Manual. e The Quality Assurance group did not review the Cuality Assurance Progrs=s of design subcontractors prior to =id-1977. t O ?G&E - March 3, 1932 5/99

4 a WYLE was not contractually required .to*have a , ,, Quality Assurance Progr1= until Dec. 1, 1979. e ANCO was not contractually required to have a Quality Assursace Progra= until May 1978.

4. PGLE design verification on. in-house activities and suppliers was unstructured and applied inconsistently.

We consider that design verification consists of the following three ele =ents:

1) Design overview for design approach, =ethods, design input selectica, and assu=ptions.

2) Detailed checking of design steps and cc=pleted design docu=ents.

3) Verifiestion of approved "As Buil:" condition against approved design. .

Activities for eier vere not initiated until 1979 and are to be revit caring Phase II. Documentation showed detailed cheers to be perfor=ed on PGaI work with design overviews .being perfor=ed on a selective basis. Most of what PGaE refers to as Design Reviews consists of ele =ent 2. e Cc=prehensive overviews and detailed checks were - performed by EDS on Class I electrical design, sc=e EVAC, and structural items. Design overviews were not evident for =echanical designs, e PG1E did not require design contractors to perfor= design reviews of their own work. e For the =ajority of cases reviewed, design verification criteria were not defined and were dependent on the discretica of the reviewing engineer. e Occu=enta ica cf design verifica:icas vas incensis-tent and 1 ti=es incc=plete.

5. There was no effective docu=en centrol systa=

estahlisned.

• Destan interfaces internally and e.:ernally were t

W-5

                                                                                    ?G&E - M. arch 3,          1982 6/39

I O . , not effectively controlled. Various organin,ations , sometimes had different revisions of, the same .. documents. - e Identifica:1cn and control of support drawings was

             ,   inadequate.            There were cases where different versicas of the same drawing revision were in use.

e Engineering groups considered the Hosgri design criteria in the FSAR as a con?. rolled docu=ent, *

• which it was not.

e There was no effective methed for controlling the Hosgri seismic data which was distributed within PG&E and to design consultants. e Historical copies of some precedures and manuals revisions were not available. e The construe:1on drawing list (January 1982) was " not accurate for sc=e of the support drawings wnich were checked. e i.pproval signatures were not entered on support

                  . drawings,        but' approvals were in the calculation package, e       There        was no evidence               that all revisions of                             -

supplier test plans, procedures and reports were . reviewed by PG&E. t O V PG&E - March 3, 1982 W-6 7fgg

O i ATTACHMENT X QA PROGRAM REVIEW REPORT - R. F. REEDY AUDIT OF WYLE l t O l

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

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QUALITY ASSURANCE REVIEW REPORT i

l PHASE I . . i i  ;

SEISMIC SAFETY-RELATED DESIGN SERVICES ,

4 2 PERFORMED FOR PG&E . l I BY l i ! WYLE LABORATORIES , 1 i PRIOR TO JUNE 1978

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Quality Assurance Program Review Report ! Phase I l Seismic Safety-Related Design Services Performed for nGLE By lll WYLE Laboratories Prior to June, 1978

                                                                                                                                                    ~

Introduction Scope: On January 14, 1982, R. F. Reedy, Inc, completed the Quality

• Assurance Review and Audit of WYLE Laboratories (WYLE).

The purpose of this review and audit was to address the adequacy of WYLE's quality assurance procedures, controls and* practices concerning the development, accuracy and transmit-tal of seismic safety-related information by WYLE to PGSE and other consultants to PGEE. The basis cf this-review and audit was to determine if the WYLE Quality Assurance Program as implemented prior to June 1978 met the applicable requirements of 10CFR50, Appendix B for the seismic safety-related design services performed for PG&E's Diablo Canyon Unit 1 Plant. , Type of Design Services: During the period of January 1974 until June 1978, WYLE performed a variety of seismic safety-related design activities which consisted of: (1) Seismic testing of 10 instruments January 1974 ' through April 1975. (2) A seismic evaluation of basic scope for Diablo Units 1 and 2 to define seismic design problem areas and to recommend seismic testing activities WYLE could perform to aid PGEE in verifying their dynamic analysis of various items of equipment. (3) Actual seismic testing of items of equipment defined under (2) above. Contracts: On January 11, 1982, the Audit Team was advised by WYLE that PGLE cont acts 5-61-77, 5-66-77 and P.O. #4R2494 were applicable for seismic safety-related services they provided to PG&E prior to June 1978. However, a review of correspondence disclosed that WYLE was requested (R. V. Bettinger letter to Mr. Drexel Smith dated September 28, 1977) to provide PG&E with their quality assuran'ce procedures X-1 NYLE - March 1, 1982 g 2/61

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d. ,

to be used for assuring WYLE's performance on safety-related

 -()              systems,     structures, or componen's under PGLE contracts 5-61. -

77, 5-66-77 and 5-67-77. The Audit Team requested a copy of contract 5-67-77 for a review to determine if its scope was for WYLE activities related to seismic safety-related design efforts by PG&E. WYLE advised the Audit Team that they had no record of such a contract number. . The Audit Team then requested that Mr. W. T. Spitzer (the PG&E observer during our audit) verify with PGLE whether or not 5-67-77 was a valid contract for WYLE seismic safety-related '. design services. Mr. Spitzer, after phone discussions advised the Audit Team that PG&E records in San Francisco indicated that contract No. 5-67-77 had not been exercised. This is considered to be an open item to be verified during the ' independent audit of PG&E for pre June 1978 seismic safety-related design activities. On the basis of the above then, the Audit Team concentrated its review on contracts 5-61-77, 5-66-77 and P.O. #4R2494. Contract 5-61-77 Contract 5-61-77 was executed between August 1977 and February - 1979 and was subdivided by WYLE into internal WYLE task numbers, ND26286, ND58255, ND58378, ND26291, ND26207, ND26301, ND26308 and ND58215. ND26286 . This WYLE task consisted of a feasibility seismic evaluation study of basic scope for Diablo Units #1 and #2 t6 define for

• PG&E problem areas in their seismic safety-related design work and made recommendations for seismic testing for verification of dynamic analysis.

ND58255 and ND58378 The actual seismic testing performed by WYLE was controlled under these two task numbers. ND26291, ND26297, ND263Ol, ND26308 and ND58215 These WYLE tasks were described as not co"ering seismic testing / design and were not considered further by the Audit Team. Contract 5-66-77 Contract 5-06-77 was executed between September 1977 and October 1977 and was given a WYLE task #ND58328. This task (:) x-2 WYLE - March 1, 1982 3/61 l l I

s . e did cover seismic testing activities for safety-related items. P.O.#4294

• g PG&E P.O. #4R294 was executed between January 1974 and April 1974 (WYLE Task #ND53744) and it consisted of seismic testing of 10 instruments. , ,

P.O. #4R294 PGLE P.O.#4R294 was executed between January 1974 and Apri1 1974 (WYLE Task #ND53744) and it consisted of seismic testing of 10 instruments. During this review, it was determined that WYLE did no.t perform any dynamic design analysis activities for PG&E prior to June 1, 1978. Evaluation Criteria The charter of the Audit Team fcr this review and au'dit of WYLE, as stated above in the scope, was to determine if the seismic safety-related design activities performed by WYLE were controlled in a manner that they complied with applicable " criteria of 10CFR50 Appendix B. Prior to this audit, the WYLE Quality Control !!anual SPP-518Q dated April 30, 1975 and Quality Control Procedures Manual SPP-518 were provided to R. F. Reedy, Inc. as the UYLE Quality Assurance Program Manual and Quality Assurance Procedures under which WYLE controlled their pre June 1978 seismic safety-related design / testing activities. These ,two manuals , were reviewed prior to performing the audit in the R. F. Reedy, Inc. offices during which several questions were developed related to the adequacy of the WYLE program to meet the applicable criteria of 10CFR50, Appendix B. These questions were given to WYLE on January 4, 1982 with a verbal request that WYLE be prepared to discuss the questions and provide to the Audit Team on January 11, 1982 answers which would be used by the Audit Team in the performance of the l implementation audit. Attached as Appendix A to this report is the Program Review Chechlist prepared for discussions with WYLE. During the discussions between the Audit Team and UYLE held on January ll, 1982 the Audit Team was not provided with adequate answers to its questions which would permit an acceptance of WYLE's Quality Assurance Program (SPP-518Q and SPP-518) as being adequate as written to meet the relevant criteria of 10CFR50, Appendix B. Attached as Appendix B to this report is the attendance list for the review and audit of WYLE. t X-3 WYLE - March 1, 19e 4/61

t i i

     ']                        Further                        discussions. and reviews of correspondence and                                                                                             '
;                                contractual files were then performed by the Audit Team in an.

effort to determine if PG&E had imposed other than SPP-518Q and SPP-518 quality assurance requirements on WYLE for pre-June 1978- seismic safety-related design activities. The - "

following are the observations of these discussions and reviews made by the Audit Team. 1 The details of this portion of the Audit Team's review are i reported in Appendix C to this report. In summary, the first l contractual requirement for quality assurance requirements were imposed on WYLE in contract change order No. 5 to cratract 5-61-77. This change required that WYLE perform -

their tests in accordance with PG&E's " SPECIFICATION FOR CONTRACTOR'S QUALITY ASSURANCE PROGRAM".

Note: The specification PG&E .actually att' ached to this

change order is titled " SPECIFICATION FOR TESTING' LAB'S QUALITY ASSURANCE PROGRAM". Change Order No. 5 required the above to be effective on December 1, 1978 and contained no provisions for re-evaluation of already completed seismic testing for compliance. .

It was the decision of the Audit Team to prepare a revised Audit Checklist (Appendix D to this report) and to Audit WYLE's imposed controls during their pre June 1978 seismic testing activities to determine if WYLE's Engineering

, Practices controlled their seismic testing activities'in a - i manner that they could be accepted as meeting the intent of - the applicable 10CFR50, Appendix B criteria. Appendix D to

• this report also contain the details of the Audit Teams Findings and Observations recorded.

It should be pointed out that the date of the earliest seismic testing WYLE performed commenced in January 1974 well after the effective date of the NRC 10CFR50, Appendix B. j criteria which was published in 1970.

Conclusions:

On the basis of our review and audit it was concluded that: (1) WYLE's Quality Control Manual SPP-518Q, April 30,

1975 and Quality Control Procedures Manual SPP-518, April 30, 1975 were not implemented in a manner whereby 3

WYLE was able to provide objective evidence of compliance with applicable criteria of 10CFR50, Appendix B, for pre June 1978 seismic testing activities. (2) Contractually, WYLE was required by'PG&E in Change s () Order No. 5 to contract 5-61-77 to perform their Quality WYLE - March 1, 1982 x_4 5/61 f i

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til Assurance in accordance with PGLE's " Specification for Contrcctor's Quality Assurance Program" as stipulated in Attachment A to Change Order No. 5 and/or in

• accordance with the " Specification for Testing LAB's Quality Assurance Program" which is the actual Attachment B to Change Order No. 5. Whichever specification is determined to be applicable for WYLE's QA program, the Audit Team was not able to verify' during the audit that WYLE did comply with either specification. The Audit Team was advised by WYLE that their Quality Assurance records applicable for the PGLE

            ,     contracts   had been destroyed without authorization

• during their QA Department's move from one building to another. As such, a meaningful audit of objective documented evidence of WYLE's QA program implementation was not possible. ,

(3) Since WYLE provided insufficient documented evidence of compliance with any Quality Assurance Program, the Audit Checklist prepared to audit for verification against WYLE's Quality Control Manual SPP-

• 518Q and Quality Control Procedures Manual SPP-518 was used to interview responsible WYLE personnel. This was to determine if WYLE's compliance to their Engineering Practices could be used to establish a QA Program equivalent to the applicable criteria of 10CFR50, Appendix B. The answers provided to the Audit Team and
• absence of documented evidence of implemented controls verify in the opinion of the Audit Team that the pre June 1978 seismic safety-related testing performed by WYLE was not performed in a manner that met the criteria of their SPP-518Q & SPP-518 Manuals and did not meet the applicable criteria of 10CFR50, Appendix B.

(4) It is the opinion of this Audit Team that the seismic safety-related testing activities performed by WYLE for Diablo Canyon cannot be accepted as having been performed and controlled under a Quality Assurance Program which met the applicable criteria of 10CFR50, Appendix B. t X-5 WYLE - March 1, 198 6/61 e

[~) -v , , Summary of Findings and Observations The primary findings of this Audit Team is that a contractual requirement that WYLE perform , t h e' seismic ..

       ....       safety-related testing activities under the controls of a

• Quality Assurance Program was not in effect prior to December 1, 1978. Insufficient objective evidence was submitted to the Audit Team to indicate that WYLE's seismic testing activities performed prior to December 1, 1978 were" in compliance with a Quality Assurance Program.

It should be verified by review of PGLE purchase records. . that Contract No. 5-67-77 was not initiated with WYLE. X-6 F)1

\ss                                                               WYLE - March 1, 1982 7/61

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O ,. .. e O QUALITY ASSURANCE REVIEW REPORT , . PHASE I ( SEISMIC SAFETY-RELATED DESIGN SERVICES 1 PERFORMED FOR PGLE By

              .                              ANCO ENGINEERS (For=erly. Applied Nucleonics, Inc.)      -

PRIOR TO JUNE'1978 . l l l i l i l i e i t lO l ANCO - March 1, 1982 1/41

ll> Quality Assurance Program Review Report - Phase I . Seismic Safety-Related Design Services Performed for PGaE By ANCO Engineers Prior to June, 1978 Introducti6n - Scope; . On J'anuary ll, 1982, R. F. Reedy, Inc. completed the Quality Assurance' Review and Audit of ANCO Engineers (ANCO). The purpose of this review and audi was to address the adequacy of ANCO's quality assurance procedures, controls and practices concerning the development, accuracy and transmittal of seismic safety-related information by ANCO . to PGLE and other consultants to PGEE. The basis of this review and audit was to deter =ine if the ANCO Quality Assurance Program as implemented prior to June 1978 met the applicable requirements of 10CFR50, Appendix 3 for the seismic safety-related design services performed for PGLE's Diablo Canyon Uni: 1 plant. Type of Design Services: . , During the :ime in question ANCO performed a feasibility study to determine if in-situ vibration testing could be performed on sele: ed seismic safety-related systems and items to verify cynamic analysis that PGLE had performed on the syste=s and items. As the result of the feasibility study ANCO performed, :ney were contracted to and did perform in-situ vibration testing on selected seismic safety-related systems and items. . C'ontrac s: CGEE had identified two PGSE General ;!anager Authorina-

lons (contracts) applicable to ANCO seismic safety-related design services prior to June 1978. These two (5-68-77 and 5-82-77) were verified by the Audit Team to be the only two pGSE contracts in effec during the time in question.

2) -entrac: 5-65 Covered a feasibill:ytstudy to determine if in-situ (Diablo Canyon Facility) tes:s could be performed on sele :ed items and systems; and if the test results would be meaningful for verification of the

       , , , . .     .-                                   Y~1 ANCO - March 1, 1982 2/41

( adequacy of PGEE's seismic designs for meeting the postulated Hosgri earthquake seismic criteria. The ANCO project / contract number for this effort was 1122.4.

                                                           ~

Contract 5-82-77: Covered the actual in-situ . b) testing ANCO performed on items and systems selected on

                                                 .             the basis of the conclusion (s) of the feasibilit) study (5-68-77).                      ANCO identified this project as 1122 4B and further subdivided it into the tasks 1122.4B.1 torough 1122.4B.10.                       Task 1122.45.5A was added in 1/79 and task                                                                                                                                                        -

1122.45.10 was performed in Auguht 1978 and Appendix therefore neither were included in this review / audit. A * ' to this report gives a breakdown listing of tne ANCO tasks related to contract 5-82-77. Evaluation Criteria: Our Quality Assurance Review and Audit of ANCO addressed the required criteria of 10CFR50, Appendix B. We evaluated the ANCO Quality Assurance Manual applicable at the time the activities were performed, reviewed

                                                         -      available documentation and records, and interviewed ANCO                                                                                                                                                           -

personnel to evaluate implementation compliance with the required criteria of 10CFR50, Appendix B. Note that ANCO Quality Assurance Implementation Procedures were not a part of this r'eview as none were prepared by ANCO for therefore ANCO's control of these activities and perfor=ance could not be evaluated against detaildd. - approyed procedures. , It should be pointed out that ANCO contracts were int:iated in 1977 and later and :ne requirements of 10CFR50, Appendix B were published in 1970. The Audit Team- could not find any evidence tha: the required criteria of 10CFR50, Appendix B, were referenced or specifically imposed on ANCO contracts. Because of this our review and audi was performed agains: ANCO's Quality Assurance Manual ANCO Spec. QAM-002, May 1978 (each of

ne ANCU Final Tes: Reports for their tas>.s 1122.43.1 through 1122.43.10 contain' a certifying statemen that the tests and reports were performed under and met the requirements of this manual). The purpose of this was determine 'wne-her the required review /auct to criteria of Appendix 3 were met.

Conclusions:

On :ne basis of our review and aud:: it was concluded

                                                                        --                                                                                                                                                t fg G                                                  (1)       The           ANCO     Quality Assurance Manual (ANCO Spec. QAM-Y-2 ANCO - March 1,                                           1982 2/41

002, May 1978) required prepara' tion, approval and implementation of a sufficient variety and number of . QA procedures whereby if accomplished _by ANCO the applicable criteria of 10CFR50, Appendix B, could* have been achieved. (2) PGLE accepted ANCO Spec. QAM-002 in April 1978 but - no documentary evidence was made available to the survey team that indicated PGhE made a for=al request that ANCO prepare- and submit to PGEE for approval the applicable Quality Assurance Implemen-tation Procedures. . - (3) ANCO did not provide documentary evidence of any actions taken to review work performed prior to April 78 for compliance to the implied' control requirements of ANCO Spec. QAM-002, May'1978. , (4) No documentary evidence was provided by ANCO to verify that activities performed from April 1978 to June 1978 yere performed in accordance with the implied control requirements of ANCO Spec..QAM-002, - May 1978. Note: ANCO advised that some QA files had inadvertently been destroyed without notification to PGEE. (5) It is the opinion of this Audit Team ths insufficient documented evidence exists, or was made available during this review / audit,.to assure and verify that ANCO, prier to June 1978, had in fact implemented a qualt y assurance program. Method of Review The method of review and audit for ANCO consisted of the following steps enumerated as (1) through (4) below: (1) An introductory' meetir.g was held with ;NCO on January 4-5, 1982. The purpose of :nts meeting was to obtain an understanding of ANCO's operatlon and the type of seismic safety-related design services they provided

c POiE. A major por;ica of :nis introdcctory meeting was consumed in an effor: to determine what ANCO Quality Assurance program Manual, practices and implementing procedures had been applied to the activities performed.

Appendix B lists the ANCO, pGLE and R. F. Reedy, Inc. personnel in attendance at all meetings with ANCO.

                                             -Y-3                  ,

O ANCO - March 1, 1992 4/41

().- Initially it was the intent of the team to discuss (2) questions which had been prepared based on a preliminary review of the ANCO Quality Assurance Manual "ANC Spec. QAM-001, December 1976" which had been provided by pG&E

             ~

to R. F. Reedy, Inc. as the controlling QAM for the pre , June 1978 Services ANCO provided to pGEE. ANCO advised the' Audit Team during these discussions that ANCO's Quality Assurance Manual "ANCO Spec. QAM-002, May 1978" was the applicable QAM for the servi'ces they performed. The Audit Team was shown copies of ANCO's Final Test Reports for their tasks 1122.43.1 through 1122.48.10 and each of these reports contained a certifying statement by that the tests reported had been controlled under ANCO and met the requirements of "ANCO Spec. QAM-002, May 1978." At this point it was necessary for the Audit Team to expand this introductory discussion and. review to establish, on the, basis of available documentary records, what ANCO Quality Assurance Program Manual, practices and procedures should be used for its review and audit of ANCO's activities. The results of these discussions and

    ,          ' reviews are reported in the detail                                the possible from the Audit . Team in evidence provided             to                                                                    .

documentary Appendix C to this report. (2) On the basis of the review performed under (2) above and reported in Appendix C it is the conclusion of the Audit Team that ANCO's Quality Assura nce Manual " ANCO , Spec. QAM-002, May 1978" applied for the ANCO activities to be audited. An Audit Checklist was prepared by the . Audit Team against this manual, and then ANCO's ,

  ,                 performance was reviewed                        and      audited   to              verify whether ANCO            did perford their activities                        under                   sufficient controls whereby                     the  Audit     Team    could   verify                  compliance I                    with. the required criteria of 10CFR50, Appendix B. This checklist and Audit Team Findings and Observations are included as Appendix D to this report.

(4) The Quality Assurance review and audit of ANCO was per:crmed on January 6 and 7 of 1982. On January 6, 1932 it became apparent that scfficient cocumentary evidence verifying the implementation of and compliance with a c quality assurance program did not exist or could not be l provided to the Audit Team waereby a determination of compliance with 10CFR50, Appendix B requirements could be verified. The audit checklist drafted for this review and audit and the Audit Team's observations and findings are incluced as Appendix D to this report. () Y-4 l ANCO - March 1, 1982 5/41 l l

af e Summary of Findings and Observations *

          . The   primary finding of this Audit Team is that a Quality                                             '

Assurance Program was not implemented by ANCO for those seismic safety-related design activities they provided to PGLE for the pre-June 1978 time period. o t Y-5 9 A:CO - March 1, 1982 6/41

, . . .._ _ _ _ . _ _ _ . _ _ = _ _ _ _ _ = . _ _ _ _ . _ . _ . - -. . _ _ . _- . -- . _ . - - . _ . _ _ _ _ _ _ _. . . . _ . - _ l 4 i-1 r i i e !O 14 1 b P i i i . t 4 > i . I 1 t r I ATTACHMENT Z QA PROGRAM REVIEW REPORT - i i R. L. REEDY AUDIT OF HARDING LAWSON ASSOCIATES I

i jf i 1 s i

i 4 ] f I f O < I i t e-w, m ee --r w c w .em. >_ -m_=ces-wm-+-- w wwm,-ww w w .-*se,---c-.. v-- - * * = + + + -

_ __ _ . _ _ _ ____ _ _ _ _ _ _ - _ _ __ _. _ _ _ _ _ _ _ _ _ _ . _ _ _ _ __~ __. __. ____ . i t l . 1 QUALITY ASSURANCE REVIEW REPORT PHASE I , , SEISMIC SAFETY-RELATED DESIGN SERVICES . , PERFORMED FOR PG&E , By , 4 HARDING LAWSON ASSOCIATES . . > ! PRIOR TO JUNE 1978 i l . . l 4 I i . 4 l . i l i

                                                                                                    -c 7

\ . I i i i ! R. T'. Reedy, Inc. j January 26, 1982 HLA 1/64 war =w = - + vg ww w e w e- ,-- p g + wre =-er + m -rg e-M e y w o-+w m- w +- e- ,wn-- v---rew-m-ww-----+,-e*n**erwn-ev- -- =---*-e*e*"~*'-em**"-**--~~ = - - ' - - - - ' - ~ " - ' " " ' ' ' - - ~ ' '

k (? ' Quality Assurance Program Review Report' .

                                                                                                 ~

Phase I Seismic Safety-Related Design Services Performed for PGSE By Harding Lawson Associates Prior to June, 1973 Introduction . Scope: On January 11, 1982, R. F. Reedy, Inc. completed the Quality Assurance Review and Audit of Harding Lawson Associates (HLA). The purpose of this review and audit was to address the adequacy of HLA's. quality assurance procedures, controls ' and practices concerning the development, accuracy and transmittal of seismic safety-related information by HLA to PGLE and other consultants to PGLE. The basis of - this review and audit was to determine if the HLA

           -Quality Assurance Program as implemented prior to June

(. 1978 met the related requirements of 10CFR50, Appendix.B for the seismic safety-related design services performed for PG&E's Diablo Canyon Unit 1 Plant. . , Type of Design Services: During the time in question- HLA performed soil investigations, geotechnical studies and consulting for the Diablo Canyon Unit 1 Plant. The principal purposes of the HLA work was to determine physical characteris-tics of subsurface caterials, provide design criteria and specifications for soil fills, provide design criteria for foundation support and to determine slope stabilities. Contracts: DGEE had identified the following series of pGEE General

               ..!anager  Authorizations (contracts) applicable to HLA prior to June 1978.      These were:

1 Z-1 HLA 1/25/32 2/64

( , , g"' (1) . (2) (3) (4) (5) GMlG7027 Letter. Authorization Contract HLA Job No. 3/14/68 64303 Eng. 11-68 569.8

• 12/10/68 64832 -

569.8

                                                                                                       ~~

12/22/69 65598 569.010.04 9/26/73 68869 569.021.04 1/15/74 68869-A 569.021.04 2/21/78 68869-B 569.031.04 5/ 3/78 68869-C 569.034.04 & 569.035.04 On January 8, 1982, we received a listing (HLA Jobs, Diablo Canyon, PGaE) of all HLA contracts with PG&E(included as Appendix B to this report). The correlation to HLA job number shown in (5), above,'is apparent and was not positively verified. Based on our review and discussion of the HLA listing with Mr. H. Taylor of HLA, we concluded that parts of each contract included seismic safety-related design work, and therefore, - our review would encompass objective evidence from each contract and none were excluded from consideration. Evaluation Criteria: s' Our Quality Assurance Review of HLA used the applicabl*e requirements of 10CFR50, Appendix B and included ANSI - Standard N45.2.20, Supplementary Quality Assurance . Requirements for Subsurface Investigations for Nuclear Power Plants," as the Quality Assurance criteria that were to be met. We evaluated the HLA Quality Assurance Manual and procedures applicable at the time the activities were performed, reviewed documentation and records, and interviewed HLA key employees to evaluate implementation compliance nith the requirements of. Appendix B and N45.2.20. It should be pointed out that HLA contracts began in 1968 and the NRC (AEC at that time) requirements of 10CFR50, Appendix B were not published until 1970. In addition, the final draft of the ANSI N45.2.20 was only issued for final balloting of the ANSI N45.2 Committee in 1978. l Further we could not find any evidence that the i requirements of Appendix B were imposed on HLA contracts. Because of this much of our review was to establish whether or not the essential criteria of these documents were met. t l ()(~ Conclusions On the basis of our review it was concluded that:

                                                  ~

HLA 1/26/S2 3/64 v e

(1) HLA was not required to implement a formal Quality y Assurance Program for their activities prior to April (( 10, 1973. (2) The HLA Quality Assurance Program and' operating . procedures applicable to the activitics performed prior to June 1978 did not prescribe adequate controls to comply with the applicable criteria requirements of 10CFR50, Appendix B, or ANSI N45.2.20.

                                                               ~

(3) There was not sufficient objective evidence available at HLA to establish that a controlled system was in

• effect which could be accepted as equivalent to 10CFR50, Appendix B.

Method of Review The method of review for HLA consisted of the following steps enumerated as (1) through (4) below; (1) An introductory meeting was held with HLA on December 28, 1981. The purpose of this meeting was to obtain an - understanding of the HLA organization, the type of design services provided to PGaE, and to verify what controlling program, practices and procedures had been applied to the work perfor=ed. We also indicated the type of Quality Assurance review we would be making. Our questions during (-. this meeting were based cn a preliminary' review of the HLA . Quality Assurance Manual and Operating Procedures. Appendix A lists the HLA, PGSE and R. F. Reedy, Inc. personnel in , attendance at all meetings with HLA. (2) We reviewed the HLA Operating Procedures and Qualicy' Assurance Manual applicable to activities performed prior to April 10, 1978. This review consisted of a comparison of the Procedures and Manual with the 10CFR50, Appendix B criteria requirements for I Organization; II Program; III Design Control; IV Procurement Document Control; V Inctructions, Procedures and Drawings; VI Document Control; VII Control of Purchased Material, Equipment and Services; XVI Corrective Action; XVII Quality Assurance Records; and XVIII Audits. These criteria were selected as being applicable to ELA based on their work. HLA also subcontracts for. certain services such as boring and laboratory testing. The Program Review checklist is. included as Appendix C of this report. The HLA Field Engineer's Notebook, October 1973, was included in the review as part of HLA's operating procedures. 1 2-3

   ,                                                          HLA    1/26/32   4/64

( llI

k 's-) - ([ (3) Because our review disclosed an inadequate Quality Assurance program in comparison to the applicable,10CFR50 Appendix B criteria, a meaningful audit checklist could - - not be used to review the HLA program. Therefore, it was ,

                                                                                              ~

decided to use the generic program review-audit checklist included as Appendiz D of this report. The basic approach was to audit what evidence did exist to determine whether or not an equivalent quality assurance system had been in - effect. (4) The Quality Assurance audit of HLA was performed on . . January 8 and 11, 1982. It became apparent on January 8 that the audit could not be conducted as structured. It was then decided to review documentation and notes and to interview for evidence of compliance. Additionally, on January 11, an interview-review against the requirements

                                                           ~

of ANSI N45.2.20, was performed. A summary of comments from the January 8 and 9 review is included as Appendix E of this report. Summary of Findings and Observations ' The primary finding [isthataQualityAssuranceprogram was not in effect for the scope of work we reviewed. f. The above findings were reviewed with Mr. Henry Taylor at (_ an exit interview and he state?. he did not disagree. . G $ e HLAs 1/26/82 .5/64 i)

                                                -x.* - -. -

d O ATTACHMENT AA QA PROGRAM REVIEW REPORT - R. F. REEDY AUDIT OF CYGNA l O

t , , . 1 i

                                                       .                                                                                                                                                     ./             .
!-                                                                                                                                                                                                                                          i.

1 l0.- ., l f I SEISMIC SAFETY-RELATED DESIGN SERVICES [ i, -. PERFORMED FOR PGEE . .. t i By ,

!                                                                                                                                                                                                                                            i i                 .

, CYGNA ENERGY SERVICES (EES)- I

{

PR'IOR TO JUNE 1974 . . i i t

l 4

l . . 1 f . 1 i i 1 . l i t - . , i 1 2 i l i l i. I i t . EIS - March 1, 1982 1 1/43 l

I Quality Assurance Program Review Report Phase I Seismic Safety-Related Design Services ' Performed for PGEE , .

         ~

By - CYGNA ENERGY SERVICES (EES) Prior to June, 1978 Introduction Scope:  ; On January 14, 1982, R. F. Reedy,.Inc. completed the - Quality Assurance Review and Audit of Cygna Energy Services (EES). The pcrpose of this review and audit was to address the adequacy of EES' quality assurance procedures, controls and practices concerning the development, accuracy and transmittal of seismic safety-related information by EES to PGaE and other consultants. to PGLE. The basis of this review and audit was to determine if the EES. Quality. . Assurance Program as implemented prior to June 1978 met the related requirements of 10CFP.50, Appendix B, for the seismic safety-related design services performed for PG&E's Diablo Canyon Unit 1 Plant. Type of Design Services: During the time in question, EES performed're-analysis of , selected piping systems, design review of piping supcor:s, analyaed Class 1 piping and re-evaluated all. C'. ass 1 3upports. Con:racts: Centrac: 5-16-77, dated March 9, 1977, was the only on:ract during this time period. The scope was to perform seismic re-analysis of selected piping systems

           .    . ac=.g..           ceci - of p.pe s.pports far PGui's Cni;s 1 anc 2,     Diablo Canyon Site, in accordance with contractor's (EES) proposal dated February 18, 1977.                                                                                                                                      EES' proposal states           ths: analysis w*                                                                                                                           "se response spectr4 for an eartaquake                occurrence postulated on :ne Hosgri tault.

Change Order Number 1, dated August 15, 1977, adced analysis of Class 1 piping and re-evaluation of suppo.~ts for all Class 1. Evaluation Criteria: - 1 Qur Quality Assurance Review of EES used ne applicable reoutrements of 10CFR50, Appendix 3, and Standarc (l) N45.2.ll as the Qualt:r Assurance Criteria to be met. We AA-1 EES M3 Ch l' 1902 2/43

                                                        *             ,                            *                                                                                                                                                 ..+

l i i

                                                                                                                                                                                                                                                      *~

(_) evaluated the EES Quality Assurance Manuals used at the time of the contract and reviewed documentation and records to evaluate implementation compliance with the Quality Assurance Program. , , Conclusions On the basis of our review of EES' Quality Assurance during this time Manuals and documentation applicable period, we believe that: - Program, EES' Quality

1. The Quality Assurance Assurance Manuals Rev. 2 (January 30, 1976) through Rev. .

4 (December 2S, 1977) in comparison to 10CFR50, Appendix B, minimally addressed the applicable criteria. ' Program implementation was inadequate as evidenced

• 2.

by the two findings and seven observations summarized as follows: Findings -

     .

• o of interface control regarding

1. There was a lack input from PGEE.

EES could not verify that one support (out of a sample of 5) sent from PGEE had ever been There was no analyzed or deleted as a requirement. - interface con:rol procedure. It is recommended that this problem be_ investigated ' to determine whether or not this was a generic problem. { -- ere not adecua:ely con:Ybiled as

2. " Controlled" memos did not indicate

• demos to content and distribution.

whether tney superseded orevious memos and were not specific in referencing other material. It is recommended that the applicable controlled conflicting ru-ss be reviewed to determine whether instructions were issuec anc eva'ua:4 pc: ::ial impar . rn the work performe'd. are included in the Audi: Finding I Soth findings Reports included in Appendix B to this report. Observations - no corrective action recuired:

1. A PGEE letter dated August 16, 1976 was referenced as being part of the project criterta by an EES me=o c..:ec 11/21/77. The referenced pGLE le::er pould not be located in the project files.

O . . . ,.

                                                                                                           ^^-2 EES - March 1, 1982 3/43 t.

2. Formal records of the computer program verification ' '

k of SAP IV were not available and the ' verification could be confirmed only with the aid of the personnel involved. . Because of the significance of the computer runs to project design it is recommended that a retrievable * . record of the verification be developed.

3. No Quality Assurance procedures other than the Quality Assurance Manual itself were developed and implemented; for example, conducting audits.

4. Quarterly management review m'eetings were not held quarterly. .

5. All revisions of Quality Assurance Manual were not formally transmitted to the Project Engineer Manager.

6. Audits were not timely or comprehensive. Only calcula:1or.s and computer binders were audited.' There was no evidence of effective corrective action.

                     . Other than the basic QA indoctrination and training session there wds no evidence cf auditor training and'                                                                  "

qualification. Method of Review The method of our review for EES consisted of the folicwing steps. enumerated as 1 trrough 4 below:

1. An introductory meeting was held with EES personnel

• __ on December 25, 1981. The purpose of this meeting was to obtain some understanding of the EIS-organization and the type of design services they performed. We also incicated the type of Quali y Assurance review we woulc be making. Our questions during this meeting were based on a preliminary review of the EES Quality Assurance "anual.

A tachmen: A lists the EES. 9GLE anc R. F. Reedy,

                          !ac.              c a r s e r. r. e '. ir          a tend.1 ce 2: all          .ee-tr. s  "t-"

_ _i..e .

2. We reviewed :ne EES Qualt:y Assurance Manual, Rev.

2, (Jan;ary CO, 107C), previded as applicable c activities for contrac: 5-16-77. This review consisted of a comparison of the Program against the 10CFR50, Appendix B, criteria requirements for (I) Organiaation; (II) Program; (!!I) Design Control; (V) Instructions, Procedures and Drawings; (VI) Document Con:rcl; (XV!) Correc-ive Ac-1,on; (XVII)

                          .i . a l '. - - Assurance Recorcs; an: ../111) Audi s.

AA-3 EES " arch 1, 19S2 h 4/43

_ ...=-- - _ _ _ . . . _ . . . . _.. . s These ,, criteria applicable to ofa Appendix design B were selected as being CTheofProgram this Review report. Checklist is included as Appendiorganizat x Checklist Where the Program Review , - question indicates a programmatic * * ' the issue omission or during the audit. was included for consideration

3. An Appendix B checklist responsive 'o audit .

t criteria and Section 6.3.1, was prepared. the referenced ANSI. Standard N45.2.ll, Checklist 4. is included as Appendix D of this reportThe completed Audi . The January Quality 13 Assurance and 14, audit of EES was performed on . evident 1982. that Quality Ascurance It immediately became i ' (September Manual Rev. 3 had 13, 1977) and Rev. 4 (December 28, 1977). June been 1978. 'pplicable during the time period prior to checklist was These were reviewed and our audit 18, 1978) was revised accordingly. Rev. 5 (April not too close included as its effectivity was effecc on the audit.to the June date to have any significant t AA-4 SES .'.! arch 1, 1982 5/4'e

O ATTACHMENT BB QA PROGRAM REVIEW REPORT - R. F. REEDY AUDIT OF BLUME O

O . QUALITY ASSURANCE REVIEW AND AUDIT REPORT -, PEASE I - By: R. F. REEDY, INC. On:URS/J.A. BLUME & ASSCCIATES, ENGINEERS SEISMIC SAFETY-RELATED DESIGN SERVICES , . PERFORMED FOR PG&E , PRIOR TO JUNE 1,1978 . i

                                                                                             ~

t . URS/Blu=e - March 5, 1982 O 1/42 l

c .

          "i k

Quality Assurance Program Review and Audit Report * ' Phase I ~' Seismic Safety-Related Design Services' Performed for PGLE By URS/J. A. Blume & Associates, Engineers Prior to June, 1978 Introduction - Scope: On January 26, 1982, R. F. Reedy, Inc. completed the Quality Assurance Review and Audit of URS/J.A. Blu=e & Associates, Engineers (Ur3/Blu=e). The purpose of this review and audit was to address the adequacy of URS/Blume'.s quality assurance procedures, controls and practices concerning the development, accuracy and transmittal of seismic safety-related , information by URS/Blume to PGEE and other consultants to' PGaE. The basis of this review and audit was to determine if the URS/Blu=e Quality Assurance Program as implemented prior to June 1978 met the applicable requirements of 10CFR50, Appendix B, for the seismic safety-related design services performed for PGLE's Diablo Canyon Unit 1 Plant.. Type of Design Services: . , During the time period in question, URS/Blume performed structural design and analysis, prepared seismic criteria, dynamic analyses of piping, reanalyses, and various l consulting services. 1 Contracts: Appendix B to this report is the listing provided to us of URS/Blu=e work performed for the pGaE Diablo Canyon Project. As is evident from this listing, most of the work performed was seismic safety-related. In addition, work was subcontracted from URS/Blu=e to others, including GEO-RECOM, Inc. (geophysical explorations and laboratory testing), Woodward-Clyde-Shepard a Associates (laboratory tests), Jason Bloom (independent review o f. *erbine building) and Wyle Laboratories. Evaluation Criteria: Our Quality Assurance Review of URS/Blu=4 used the-BB-1 URS/31ume - March 5, 1. g. 2/42

O 10CFR50, Appendix B, and applicable _ requirements of Standard N45.2.ll as the Quality Assurance Criteria to be met. We evaluated the URS/Blu=e Quality Assurance Manuals used during the time the activities were perforged and , reviewed documentation and records to evaluate implementa- . . . _ tion compliance with the Quality Assurance Program. Conclusions . On the basis of our review of URS/Blume's Manuals and documentation applicable during this time period (pre-June, 1978), we conclude that: establish or implement a Qhality

1. URS/Blume did not Assurance program that met the applicable re-quirements of 10CFR50, Appendix B. ,.

2. There was no objective evidence that an equivalent program or system of controls was in effect during this time period.

Findings

1. URS/Blume did not establish or implement a Quality Assurance program that met the applicable require-ments of 10CFR50, Appendix B. -

2. The Hosgri Report was not developed or issued by URS/Blume as a controlled design docu=en..

Observations - The results of the URS/Blume review were categorized as comments rather than Observations because no QA program was used for this work. Selected comments noted during the review are included in Appendix E to this report. Method of Review The method of our review for URS/Blu=e consisted of the following steps enu=erated as 1 through 5 below:

1. An intreductory meeting was held with URS/Blu=e personnel on Decemoer 28, 1981. The purpose of this meeting was to obtain some understanding of the URS/ Bit =e organization and the type of design services they perfor=ed. We also indicated the type of Quality Assurance review we would be making. Our questions during this =eeting were based on 1

() BB-2 URS/Blume - March 5, 1982 3/42

preliminary review of the URS/Blume Quality Assurance

     ~

Manual. Attachment A lists the URS/Blume, PGLE and R. F. Reedy, Inc. personnel in attendance at all meetings with URS/Blume. . .

2. We reviewed the URS/Blume Quality Assurance Manual, Rev. 2, (November 19, 1976), provided as applicable to URS/Blume activities prior to June 1978. This review consisted of a comparison of the Program against the 10CFR50, Appendix B, criteria re-quirements for (I) Organinstion; (II) Program; (III)

Design Control; (IV) Procurement Document Control; (V) Instructions, Procedures and Drawings: (VI) . Document Control; (VII) Control of Purchased Material, Ecuipment, and Services; (XVI) Corrective Action; (IVII) Quality Assurance Records; and (XVIII) Audits. These criteria of Appendix B were. selected as being applicable to a design organination such as URS/Blume. The Program Review Checklist is included as Appendix C of this report. Where the Program Review Checklist indicates a programmatic omission or questien the issue was included for consideration during the audit. (Subsequent to the Manual review - we were informed that the Quality Assurance Manual was not invoked in PGEE orders prior to June, 1978.)

3. An audit checklist derived from the URS/Blu=e QA Manual and which was responsive to the referenced Appendix B criteria- and ANSI Standard N45.2.11, v e ction--Gr3 cl , -was prepared. During the audit (4., .

below) it became evident that our audit checklist was , not applicable and it was not used; and, therefore it is not included in this report.

4. The Quality Assurance audit of URS/Blume was l perfor=ed on January 18 and 19, 1982, with a follow-up visit on January 26, 1982. During review of the listing of work perfor=ed by URS/Blume for PGuE i

(Appendix B to this report) it was stated that a partial URS/Blume quality assurance program had been applied only. for contracts *: 6907-27(0902) In-vestigation and Design of Turbine Building for New 7.5 =agnitude Hosgri, started July 1976; 6902-2S Diablo Piping Reanalysis, started March 1977; 6902-29 Diablo Unit 2 Hanger Review, started October 1977; and, 6902-30 Diablo G-Line Anchor Review. URS/Blume stated that the partial program was defined as that outlined in an August 30, 1977, letter DAL to JOS. (Included in Appendix D to this report.) Prior l

• Shown with a single asterisk in Appendix'B to this report.

BB-3 URS/Blume - March 5, 1se_ 4/42

() e - to that time, August 1977, Blume stated that an informal program of " good engineering practices" had been used. Blu=e further stated that the , Quality , Manual was not applied to PGaE work in its entirety .- until June 1, 1978. Blume informed us that the

• Quality, Assurance program was not invoked unless required by the client.

PG&E letter, FFM to JAB July 11, 1978, required all further work on the Hosgri seismic evaluation to be conducted in accordance with the URS/Blume Quality Assurance Manual dated December 16, 1977. Note that ' this is QAM Revision 4, which was approved by PG&E in the letter JP to RFR, March 1, 1978.

                .              At this point we determined that for the Phase One review,     an adequate formal Quality Assurance program had not been established for PGLE work performed by URS/Blume.        We then attempted to review existing objective evidence against our audit checklist to determine       whether            or not there was documented evidence that " good engineering practices" were used                                         .

which could be evaluated as an informal equivalent controlled system. The objective evidence reviewed 1 did not support such an equivalency having been in j effect prior to June, 1978. There was no evidence of design inputs or. design documents being controlled or of design verifica*. ion being performed.

5. On January 26, 1982 we revisited URS/Blume to ,

determine whether or not the work perfor=ed prior to June, 1978~ was reverified after that date under the provisions of an acceptable Quality Assurance program. If URS/Blume had rever.ified the Hosgri evaluation under the provisions of an acceptable QA program this would provide a chance to minimize the impact of the design work performed prior to June 1978. During this visit it was determined that Blume did not do a complete reevaluation of the Hosgri b i design work under this later version of the QA ' program. 1 {T

   '~[,

BB-4 URS/Blume - March 5, 1982 5/42

O i ATTACHMENT CC QA PROGRAM REVIEW REPORT - R. F. REEDY AUDIT OF EDS NUCLEAR O

O Q Los Gatos, California 95030 * (408):154-9110 e

• January 20, 1982 . .

Robert L. Cloud and Associates, Inc. 125 University Avenue Berkeley, CA 94710

Subject:

Report of R. F. Reedy, Inc.

• Review of EDS Nuclear, Inc.

Dear Mr. Cloud:

Attached is a of our Quality Assurance Review leport of EDS Nuclear, copy . Inc. (EDS). We have reviewed the EDS Quality Assurance Program as it related to seismic safety-related design services performed for Pacific Gas and Electric Company prior to June, 1978. All details

• of our review are contained in this final report.

4 Ve - uly yours, Ch Roge k#4. . Ree..y, P.E., R. F. REEDY, INC. RFR:na Encl. . .

                                                                             @@[?00$
                                                                        .s                         ; .

i\ Je! 2 51982 4 i L\ M9 L L_ t CC-1 ()) Itoger l'. Rceily. I'.E. - linginscrin;! C .n ..tiing

Quality Assurance Program Review Report Phase I Seismic Safety-Related Design Services Performed for PG&E . - By ' EDS Nuclear, Inc. Prior to June, 1978 Introduction Scope: On January 6, 1982, R. F. Reedy, Inc. completed the Qualit'y Assurance Review and Audit of EDS Nuclear, Inc. (EDS). The purpose of this review and audit was to a'ddress the adequacy of EDS' quality assurance procedures, controls and practices concerning the development, accuracy and trans-mittal of seismic safety-related information by EDS to PGSE and other consultants to PGSE. The basis of this review and audit was to determine if the EDS Quality Assurance Program as implemented prior to June 1978 met the related requirements of 10CFR50, Appendix B for the seismic safety-related design services performed for PGSE's Diablo Canyon Unit 1 Plant. Type of Design Services: During the time in question, EDS performed a. structural . evaluation of pipe anchors for piping systems installed at the Diablo Canyon Unit 1 Plant. The purpose of the evaluation was to review previously designed pipe anchors for new design conditions. Contracts: On December 22, 1981, PGSE identified three contracts which had been made with EDS prior to June 1978. These contracts were numbered 5-25-71, 5-25-74, and 5-65-77. On January 5, 1982, we received a listing (Summary of EDS Experience with Pacific-Gas and Electric Company) of all EDS contracts with PGEE. The above-listed contracts were included, along with all other contracts between EDS and PG&E to date. Since all job tasks for these contracts were sequentially numbered, it was easy to verify the PGSE contract list. We reviewed the scope of con tract 5-25-71, which pertained to the dead load and thermah analysis of a pipeline. No seismic safety-related work was performed and g therefore this contract was not made part of our review. W CC-2 EDS - January 20, 2/38

O PGLE contract 5-5-74 was also investigated. The scope statement in the contract 'vas not clear about whether or not the work involved was seismic safety-related. However, a review of the documented work performed by EDS and the final EDS report showed that the job consisted of a design review' - of the . containment insolation system, the Class lE Electri-cal System _for_ the control room pressurization system and the HVAC. The ~ design review was made to determine if the pGSE design input considered all available data and if Regulatory Guides and other codes and standards were properly considered. No calculations or analysis of any seismic safety-related equipment was performed by EDS. Therefore this contract was not subject to our review.

• Contract 5-65-77 was for the structural evaluation of pipe anchors for four load conditions '(Normal, Design Earthquake, Double Design Earthquake, and Hosgri Earthquake), including calculations of design earthquake deflections and pipe anchor stiffness coefficients. The deflections and pipe anchor stiffness coefficients. The EDS Job Numbers were 0170-007 for Unit 1 and 0171-008 for Unit 2. In general, we restricted our review to Unit 1 work, although some Unit 2 '

documentation was reviewed because it was filed with Unit 1 documentation.- Evaluation Criteria: Our Quality Assurance Review of EDS used the applicable , requirements of 10CFR50, Appendix B and Standard ,N45. 2.11 as , the Quality Assurance Criteria to be met. We evaluated the EDS Quality Assurance Manual used at the time of the contract and reviewed documentation and records to evaluate implementation compliance with the Quality Assurance Program. . Conclusions On the basis of our review of the EDS Quality Assurance Manual, records and design documentation, we feel that EDS successfully implemented a Quality Assurance program which met the applicable requirements of 10CFR50, Appendix B and ANSI N45.2.ll. We have no Quality Assurance findings which require corrective action to be taken by EDS. Method of Review . The method of review for EDS consisted of the following steps enumerated as (1) through (4) below: , \- (1) An introductory meeting was held with EDS Quality CC-3 Eba - January 20, : 3/38

Assurance personnel on. December 22, 1981. The purpose , of this meeting was to obtain some understanding of the llh EDS organization and the type of design services they perform. We also indicated the type of Quality Assurance review we would be making. Our. questions during this meeting were based on a preliminary review ' of the EDS Quality Assurance f!anual. Attachment A lists the EDS, PGEE and R. F. Reedy, Inc. personnel in attendance at all meetings with EDS. (2) We reviewed the EDS Quality Assurance Manual, Revision 11, applicable to activities for contract 5 577. This review consisted of a comparison of the, Program against the 10CFR50, Appendix B criteria requirements for (I) Organization; (II) Program; (III) Design- Control; (V) Instructions, Procedures, and Drawings: (VI) Document Control; (VII.) Corrective Action; and (XVIII) Audits. These criteria of App.endix B were selected as being applicable to a design organization such as EDS. The Program Review Checklist is included as Appendix B of this report. Where the Program, Review Checklist indicates a programmatic omission or question the issue was included for consideration during the audit. (3) Audit checklists responsive to the referenced Appendix B criteria and ANSI Standard N45.2.ll, Section 6.3.1, were prepared. Criterion VI, Document Control, was not included because it was not directly . appropriate to a historical audit to be performed. . Document control was indirectly verified by review of documents and the other criteria. The Audit Checklist is included as Appendix C. (4) The Quality Assurance audit of EDS was performed on January 5 and 6, 1982. Five observations (which do not require corrective action) were noted as given in (a) through (e) below. (a) We could not verify the existence of the " Design Review Criteria" used in accordance with QAP 3.7, paragraph 3.1 for the July 26, 1978 Design Review. Evidence exists that this procedure was followed, but the documentation from 1977 was not saved. Therefore it is not possible to dir ectly verify that the " Design Review Criteria" dccument met all the EDS requirements for the July 26, 1978 Design Review. (b) In many cases, " point of use" references were not made for equations in the ggg calculation files. These CC-4 EDS - January 20, lE 4/38

A

 -b s_/                                                                                                    ,,

references were required by QAP 3.4, paragraph 3.2.1 of the present EDS Quality Assurance Manual. (c) We could not verify that the Quality Assura'nce Manager .- concurred with the resolution of discrepancies as required by QAP 3.7, paragraph 3.5, because there were - no sign-offs on the documentation. (d) We were unable to verify that project personnel had read the project interface instructions as required, but did not find any interface problems of significance. * - (e) Memoranda - were used in lieu of technical instructions. This is not consistent with the EDS Quality Assurance Program. The Program Review Checklist of Appendix B addresses a comparison of the Quality Assurance Manual and procedures-with the requirements of 10CFR50, Appendix B. Where the checklist indicated a problem of the Manual or procedures

• meeting the requirements of 10CFR50, Appendix B, a further and more detailed investigation was addressed during the implementation review. No significant items of concern were found to exist. The observations listed above address all aspects of our complete review.

l j . i t

     )                                                      CC-5

(".J x EDS - January 20, 1: 5/3S l

O ATTACHMENT DD PGSE TESTIMONY REGARDING QA/QC OCTOBER, 1977 ASLB HEARINGS l l O

l l l I O ss ,, 1 TESTIMONY OF RUSSELL P. WISCHOW ON BEHALF OF *

• 2 ,,

PACIFIC GAS AND ELECTRIC COMPANY 3 DOCKET NOS. 50-275, 50-323 - 4 BACKGROUND INFORMATION 5 A construction permit for the Diablo Canyon Nuclear 6 Unit 1 was issued by the Atomic Energy Commission (AEC) on . l I 7 April 23, 1968. i As design and construction of,the plant 8 progressed, quality assurance criteria were being f.ormulated 9 by the AEC and on April 17, 1969, the proposed " Quality Assurance 10 Criteria for Nuclear Power Plants" (or "18 Criteria") were . 11 issued. On November 1, 1969, PGandE organized its Quality 12 Engineering Section (now called Quality Assurance Department) 13 to be responsible for assuring Management on the quality of 14 design and construction of nuclear power plants,. , 15 PGandE has continually studied and appropriately 16 modified the quality assurance program and organization to be 17 responsive to changes in Nuclear Regulatory Commission Regula-18 tions, guidelines published by the Nuclear Regulatory Commission, 19 and new industrial standards,which have been accepted by the 20 Nuclear Regulatory Commission and incorporated into applicable 21 NRC Regulatory Guides. 22 PGandL's cuality Assurance Manual for Diablo Canyon 23 Nuclear Unit 2 was issued in January, 1970. PGandE a.gr.eed that 24 this manual would also be used for Nuclear U. nit I considering the () 25 status of design and construction at that time. The Atomic 26 Energy Commission issued Appendix B to Title 10, Code of DD-1

O 1 Federal Regulations, " Quality Assurance Criteria for Nuclear 2 Power Plants" on Jun,e ll7, 1970.

                                                                  ~

On December 9, 1970, the A$C 3 issued th6 construction permit for Diablo Canyon Nuclear Unit 2. 4 On March 1, 1972, the Quality Engineering Section was 5 redesignated the Quality Assurance Department and instead of 6 reporting to the Vice President-Engineering, the Qualit 7 Assurance Department thereafter reported to the Senior Vice 8 President (Engineering, Construction, and Planning and Research) 9 who is now the Executive Vice President (see below). Respon-10 sibilities of the Quality Assurance group remained unchanged. 11 pGandE issued the Quality Assurance Manual for 12 Operating Nuclear Power Plants on September 26, 1975. This 13 Manual describes the Quality Assurance Program implemented at 14 the Humboldt Bay Power Plant and the program which is being 15 implemented at the Diablo Canyon Nuclear Units 1 and 2 upon 16 the transfer of the responsibilities for the systems or plants 17 to PGandE Division Operations. 18 ORGANIZATION 19 General 20 An organization chart which delineates the PGandE 21 quality assurance activities applicable to nuclear plants is shown in Attachment A. Management Organization '24 The President and Chief Executivq Officer, hereinafter 25 referred to as the President, has the overall management respan O sibility for all nuclear plants. DD A (f *' 1 The Executive Vice President who is responsible for 2 Electric Operations, Materials, and Quality As'surance Departmeryts o, 3 (hereinafter referred to as the Executive Vice President) -

      ,'4  holds overall management responsibility for development of 5   the quality assurance program in accordance with Company policies 6  and regulatory requirements. He directs the activities of the 7  Quality Arsurance oepartment and receives and reviews audits 8  prepared by the Ouality Assurance Department.

9 The Senior Vice President (Engineering, General 10 Construction, Planning and Research, and Environmental Quality), 11 hereinafter referred to as the Senior Vice President, through 12 departments reporting to him, provides technical support as 13 necessary to the Electric Operations Department for plant 14 modifications, construction, and other engineering related - 15 activities. Those individuals reporting directly to the Senior 16 Vice President are the Vice President, Planning and Research, 17 Vice President, Engineering, Vice President, General Construction, 18 and Director, Environmental Quality. 19 The Vice President, Electric Operations reports to the 20 ' Executive Vice President and provides functional direction, 21 guidance, and assistance concerning plant safety, operation, 22 maintenance, and engineering. 23 The Vice President, Division Operations is responsible 24 for the administration of Nuclear Units 1 and 2 at Diablo Canyon. s 25 The Director, Quality Assurance reports to the 26 Executive Vice President and is responsible for the direction D D h 1 and management of the Quality Assurance Program. 2 QUALITY ASSURANCE DEPARTMENT ' 3 - The Quality Assurance Department is organizationally 4 independent of design and engineering, construction, modifica-5 tioi , materials procurement and handling. and power plant 6 operation and has the authority and organizational freedom to 7 investigate any safety-related nuclear activity, initiate 8 action which results in solutions, and verify implementation 9 of solutions to those problems involving m ity pertaining to 10 any PGandE department. The Quality Assurance Department has 11 14 technical people assigned to the Diablo Canyon Project,,7ou s 12 of whom are on site. ' 1 13 Should there be a breach of any part of the Quality 14 Assurance Program or the technical or regulatory requirements 15 at the plant where public health or safety could be involved 16 and suitable amendatory actions are not made by the responsible 17 organization, the Director, Quality Assurance, has the responsi-18 bility and authoricy to st)p the worke I'f stopping the work 19 would involve changing power level or separating a generating 20 unit from the PGandE system concurrence of the Executive Vice 21 President is required. In these instances or in the absence of 22 the Director, Quality Assurance, the Executive Vice President 23 will make the decision as to the actior, to be taken. The 24 Director, Quality Assurance is a member of ghe President's

• 25 Nuclear Advisory Committee, the General office Nuclear Plant 26 Review and Audit Committee, and is Chairman of the General Office DD 1

i g ( ,, l 1 Review Group for matters concerning reports to the NRC pursuant *- ~ 2 to 10 CFR 21. 3 QUALITY ASSURANCE PROGRAM 4 The purpose of the Company Quality Assurance Program 5 is to provide the requisite controls over those structures, 6 systems, components, and activities associated with a nu'elear 7 power plant'that are considered to be safety-related and 8 which are involved with the planning, desigh, procurement, 9 construction, startup, operation, maintenance, repair, modifi-10 cation, and inspection, all of which also includes the training 11 of personnel. The program takes into account the need for 12 special controls, processes, test equipment, tools, skills, 13 and training and provides control over work which is performed 14 by Company Departments, suppliers and contractors. Company ' 15 Departments, suppliers and contractors performing work on the 16 PGands nuclear plants or any nuclear activity related thereto 17 are required to institute and maintain quality assurance pro-18 grams approved by PGandE. PGandE performs inspections of 19 Company Departments, suppliers, and contractors activities by 20 use of Company inspectors and consultants who were not directly 21 involved with the activity being inspected. 22 This program is described in volumes 1 and 2, Quality 23 Assurance Manual, Nuclear Unit 2., Diablo Canyon site. Volume 1, 24 Policy, describes the quality assurance program in terms of 25

   )       PGandE's basic and supplementary policies for quality assurance.

26 , Volume 2, Quality Assurance Procedures, contains written DD e 1 procedures implementing the policies stated in Vol.ume 1. 2 The program for design, construction, and startup bf 3 all Design Class I structures, systems, and components requires 4 that all Company Departments, contractors and suppliers estab-5 lish and maintain in effect quality assurance programs appro-priate to the importance of their activities to safety. Requirements for contractors' and suppliers' quality assurance 8 programs are prescribed in design specifications. Specified 9 requirements are based on Appendix B, 10 CFR 50. Contractors 10 and suppliers are not permitted to proceed with their work 11 until they have submitted a Quality Assurance Manual describing 12 their quality assurance program and received approval from 13 PGandE. 14 The major design activities were shared by PGandE , 15 wt.o performed the major design work, and Westinghouse, who 16 supplied the Nuclear Steam Supply System (NSSS). PGandE con-17 ' tracted for some design work, other than the NSSS, with certain 18 suppliers and contractors. Plant construction was done by 19 contractors, with PGandE administering the contracts and 20 performing field inspections and audits. Preoperational and 21 startup testing of the plant is being performed by PGandE. 22 The Company's cuality assurance program for plant 23 operation complies with 10 CFR 50, Appendix B and with the 24 guidance provided by the following documents and the Regulatory 25 Guides and Standards referenced therein where appropriate: O W 26

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DD-6

A L,) 1 WASH 1284, October 26, 1973, " Guidance on Quality '- Assurance Requirements During the 2 Operations Phase of Nuclear Power Plant's" . 3 WASH 1309, May 10, 1974, "Guldance on Quality Assurance. Requirements During the 4 Construction Phase of Nuclear Power Plants" 5[ WASH 1283, May 24, 1974, " Guidance on Quality 6 Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants - Revision 1" 7 8 The guidance provided by the ANSI approved design 9 and construction related standards which are referenced in 10 these documents are applied to those activities which will t . 11 occur during the operational p ase that are comparable to 12 activities occurring during initial plant design and construction. 13 The PGandE Quality Assurance Manuals are approved by 14 the President, controlled by the Director, Quality Assurance,

• 15 and contain the Company quality assurance policies and programs.

16 These manuals are the governing documents for Company quality

       '17    assurance procedures. The Director, Quality Assurcnce, is 18 responsible for the preparation, revision, review, approval, 19 and distribution of these quality assurance manuals and also 20    the Quality Assurance Department Procedures Manual.         Revisions 21 to procedures shall be made as often as is deemed appropriate 22 to describe the current program; suggestions for revisions are 23 encouraged from all affected departments. The Director, Quality 24 Assurance, maintains a controlled list of those persons who have 25 manuals assigned to them and is responsible for issuing changes 26 and additions as necessary to keep the manuals in an up-to-date 27    condition.

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RESUME {^\

                               R4 P. Wischow, Director                                   * '

i Quality Assurance Department Pacific Gas and F.lectric Company 77 Beale St., San Francisco, CA 94106 , , Education B.S. Chemistry, North Dakota State University, 1951 M.S. Chemistry, North Dakota-State University, 1952 Ph.D. Chemistry, Vanderbilt University, 1958 Experience . . 9/76 to Present: Pacific Gas and Miectric Company, Director, Quality Assurance. Responsibic for directing and implementing the corporate quality assurance program for nuclear activities. Directs the Genc.al Office and site activities to verify compliance by the Company with the applicable Nuclear Regulatory Commission rules and regulations and the referenced requirements such as guides and standards.

                                                                                               ~

3/70 to 9/76: E. R. Johnson Associates, Inc., Washington, D.C. area (Vienna, Virginia) and Nuclear Audit and Testing Co., Inc., Vice President and President, respectively. Primary responsibilities were to develop, coordinate, and perform programs for: (a) performing corporate quality assurance audits for nuclear utility companies; (b) nuclear fuel reprocessing surveillance, bid evaluations, and preparation of bid specifications; (c) nuclear fuel fabrication . surveillance, audit, and inspection; (d) source inspection of , component materials for nuclear fuel manufacturers; (e) inspection and representation during UF6 withdrawal at the diffusion plants; (f) surveillance during transfer of plutonium at and between locations; (g) preparation of safeguards programs; and (h) prepara-tion of a Safety Analysis Report for a major reprocessing plant. These duties involved coordination and contact with companies throughout the world and with particular emphasis on the domestic nuclear companics. i ! 9/67 to 3/70: U. S. Atomic Energy Commission, Washington, D.C., Director, Division of Nuclear Materials Safeguards. Responsible for developing the division when it was first formed in 1967. .Estab-lished, developed, administered, and enforced the AEC nuclear materials safeguards licensing program and coordinated the detection methods and procedures for control of special nuclear materials. Maintained staff for field surveillance over nuclear materials; l approved ' safeguards programs of AEC licensees; asristed in the ! development, testing, evaluation of new instruments and plant safeguards procedures; directed staff studies on safeguards programs; and consulted with the staff of the International Atomic Laergy (~') Agency and with other countries on the requirements of domestic safeguards and accountability programs. DD-9

s Page 2 Resume - R. P. Wischow Experience (continued) 10/65 to 9/67: Nuclear Fuel Services, Inc., Washington, D.C., , , Assistant General Manager, West "a l l e.y RepmcesaiJ te Plan t. Responsible .,. for the operation and managem.ent of the Technical Services, Analytical Chemistry, Health and Safety, Licensing, and Research and Development Departments. Directed preparation of procedures which were used by the LAEA for its first inspection of a spent fuel reprocessing plant. 4/63 to 10/65: Martin Company, Nuc1 car Division, Baltimore, Maryland, Supervisor, Chemistry Unit. Responsible for supervision and coordina-tion of the chemistry effort in the Nuclear Division. Involved . synthesis of compounds, measurement of physical properties, fuel forming procedures, materials compatibility, nuclear safety, re-entry analyses, and engineering studies. Major effort was directed toward utilization of isotopes as heat sources in space and terrestrial electric systems, analytical services for reactor fuel element production and reactor chemistry. Supervised inspection and audits of reactor construction and operations at the PM-1, PM-3A, and MH-1A sitcs. ' 6/61 to 4/63: Callery Chemical Company, Callery, Pennsylvania, Senior Rescarch Chemist. Research on high energy oxidizers and propellents with emphasis on nitrogen-fluorine chemistry. 6/52 to 6/61: Union Carbide Nuclear Company, Oak Ridge, Tennessee, Chemical Technology Division. 6/59 to 6/61: Chemist-Group Leader, responsible for the , development of techniques for'the recovery of fission products from reactor fuel wastes; work resulted in patented process for Sr-90 recovery. 8/58 to 6/59: Chemist, Staf f Member, Oak Ridge School of Reactor Technology. 12/55 to 8/58: On scholastic leave of absence to Vanderbilt. 6/52 to 12/55: Croup Leader in charge of the development of the Thorex Process to recover thorium, uranium, and protactinium from irradiated thorium fuel. Joint patent holder on Thorex Process. Societies . American Nuclear Society; American Chemical Society; Fellow, American Institute of Chemists; Sigma Xi; American Society g for Quality Control DD-10

Resume - R. P. Wischow Page 3 0 Related Activitics

1. Member, Executive Committee, Atomic Industrial Forum Safeguards sa committee (1971 to 1976) . .

2. Member, Atomic Industrial Forum Subcommittee on Isotope Utilization (1965 to 1967)

3. Member, Executive Board, Institute of Nuclear Materials Management (1971 to 1973)

4. Co-Chairman ISAEC Safeguards Advisory Committee (1967 to 1970) , ,

5. Member-at-large, Atomic Industrial Forum Safeguards Committee (1970 to 1976) .

i 9 ( l l DD- 11 A(<

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