ML20209E951

Forwards Revised Response to Technical Info Request 7 Re Qa/Qc Procedures & Response to Technical Info Request 11 Concerning Conservatisms within Alternate Analysis Criteria

ML20209E951
Person / Time
Site:	Sequoyah
Issue date:	09/04/1986
From:	Gridley R TENNESSEE VALLEY AUTHORITY
To:	Youngblood B Office of Nuclear Reactor Regulation
References
NUDOCS 8609110401
Download: ML20209E951 (14)

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_e TENNESSEE VALLEY AUTHORITY CHATTANOOGA, TENNESSEE 37401 SN 157B Lookout Place SEP 0 4 S86 Director of Nuclear Reactor Regulation Attention: Mr. B. Youngblood, Project Director PWR Project Directorate No. 4 Division of Pressurized Water Reactors (PWR)

Licensing A U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Youngblood:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 By my August 18, 1986 letter to you, we provided Interim Acceptance criteria for our Sequoyah Nuclear Plant that was developed to be used for the temporary resolution of several critical engineering issues. In addition, responses were provided to 11 items of a technical information request (TIR) from NRC in a July 18, 1986 meeting. As stated in the August 18, 1986 letter, changes to the responses were to be provided as a result of your August 7, 1986 letter to S. A. White.

Enclosure 1 provides a revision to the TIR item No. 7 response in the August 18, 1986 letter regarding quality assurance and quality control procedures. Enclosure 2 provides a response to item No. 11 of your i

August 7 1986 letter regarding conservatisms within alternate analysis

criteria. This issue was not discussed in any of the TIR responses provided in the August 18, 1986 letter. The remaining TIR responses provided by my

, August 18, 1986 letter are not revised as a result of the minor variations in the questions provided by your August 7, 1986 letter. Rnclosure 3 provides the pages, one through five, that were inadvertently omitted because of a reproduction error from the enclosure to my August 18, 1986 letter to you.

I If you have any questions regarding this subject, please get in touch with M. R. Harding at (615) 870-6422.

Very truly yours, TENNESSEE VALLEY AUTHORITY

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R. L. Gridley Director Nuclear Safe y and Licensing Enclosures cc: See Page 2 00 8609110401 860904 i PDR ADOCK 05000327 g 8{t P PDR An Equal Opportunity Employer

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i Director of Nuclear Reactor Regulation SEP 0 4 886 cc (Enclosures):

U.S. Nuclear Regulatory Commission Region II Attn: Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW Suite 2900 Atlanta, Georgia 30323 Mr. Carl Stahle, Sequoyah Project Manager U.S. Nuclear Regulatory Comission

7920 Norfolk Avenue Bethesda, Maryland 20814 5

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ENCLOSURE 1 Revised Response to TIR Item Number 7 Submitted in August 18, 1986 Letter to B. J. Youngblood NRC Question No. 7 What are the QA/QC procedures that will be applied to Phases I and II of the restart program? Describe any differences between these procedures and those used in the original cable tray design and construction.

TVA Responses THE QA/QC procedures that presently exist and that will be applied to Phases I and II of the restart program are defined in Table A (attached).

The principal documents that define QA reg'11rements are: the " Nuclear Quality Assurance Manual," the " Division of Nuclear Engineering Procedures Manual," and the "Sequoyah Project Manual."

The principal documents that provide QC requirements are: the " Nuclear Quality Assurance Manual" and the " Modifications and Additions Instruc-tions."

The QA/QC procedures that were utilized during the original design and construction of the cable tray supports are also defined in Table A. The principal documents that defined QA requirements were: the "SQN Quality Assurance Manual,' the " Office of Engineering Design and Construction (OEDC) QA Manual," the "0EDC Program Requirements Manual " the " Inter-divislanal Quality Assurance Procedures," and the " Division of Engineering Design Engineering Procedures." It should be noted that design control procedures were contained in all the above documents with the exception of the "0EDC Program Requirements Manual" and the "Interdivisional Quality Assurance Procedures".

The principal documents that defined QC requirements for construction were the design drawings, construction specifications and procedures, and inspection instructions.

The problems that have been identified in the QA/QC area associated with cable tray supports can be classified generally as design related or related to configuration control. Problems were due primarily to fragmented implementation of design control associated with seismic design of cable tray supports. Other identified problems can be related to design errors and the evolution of standard practice related to the design of support baseplates.

In the area of configuration control, the problems that have been identified for cable tray supports can be attributed primarily to a failure to properly identify, evaluate, and document "as constructed" conditions that affected the design of supports. However, TVA has made, and is

making, several changes in design and configuration control that will avoid recurrence of these problems in the future. These changes are described below:

Design and Configuration Control Program A comprehensive program has been implemented to correct the problems caused by the past design and modifications control system. A strengthened engineering organization and procedural system ensures that quality engineering work is performed based on the actual configuration of the plant and proper evaluation of safety issues. It provides for reestablishing the plant design basis and establishing a configuration control system that facilitates appropriate engineering input for modifications.

A transitional design control system will be implemented that improves the existing design control process. It is based on modifying the existing TVA control system, facilitating a transition to the new permanent TVA system and providing comprehensive design packages. Design changes to the permanent plant will utilize a unitized integrated design package.

The design change package is reviewed to ensure that a quality engineered modification has been prepared; that it can be installed and tested; and that documentation is traceable, updated, and represents the actual plant configuration.

A responsible task engineer will be assigned to ensure the design work is performed; ensure the package is assembled: assist in the implementation; and ensure the work is tracked, completed, and closed.

Design changes will be developed on a unitized basis and will be of manageable size to implement in a defined period. The USQD review will be representative of the completed design package and updated if necessary after installation and testing to reflect the final installed conditions.

TABLE A ,

QA/QC Procedures A. Design and Construction QA/QC Procedures for Phase I (before restart) and Phase II (after restart)

Design O Nuclear Quality Assurance Manual O Division of Nuclear Engineering Procedures Manual O SQN Project Manual O Branch Procedures O Cable Tray Support Design - Sequoyah Nuclear Plant - Design Criteria for Category I Cable Tray Support Systems - Design Criteria SQN-DC-V-1.3.4.

O Steel Containment Vessel Design - Tennessee Valley Authority Design Specification SNP-DS-1705-9803 Sequoyah Nuclear Plant Units 1 and 2 Structural Steel Containment Vessels for the Reactor Buildings.

O Civil Design Standard DS-C1.7.1 - General Anchorage to Concrete Construction O

! Nuclear Quality Assurance Manual O Sequoyah Nuclear Plant - Specification N2G-877 for Identification of Structures, Systems, and Components Covered by the Sequoyah.

O Nuclear Plant Quality Assurance Program.

O Modifications and Additions Instructions O Design Drawings 4

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B. Original Design and Construction OA/OC Procedures pesign O SQN Quality Assurance Manual O

OEDC QA Manual-0 OEDC Program Requirements Manual 0 Interdivisional Quality Assurance Procedures O Division of Engineering Design Engineering Procedures O Cable Tray Support Design: Sequoyah Nuclear Plant - Design Criteria for Miscellaneous steel components for Class I

Structures - SQN-DC-V-1.3.2. ,

a O Steel Containment Vessel Design - TVA Design Specification

! SNP-DS-1705-9803-02 O Branch & Project Procedures Construction 1

O SQN Quality Assurance Manual O

OEDC QA Manual 0 Interdivisional Quality Assurance Procedures I O Construction Quality Assurance / Quality Control: Drawings required certification that materials conform to appropriate ASTM specification and that fabrication and erection be performed in accordance with AISC.

O Sequoyah Nuclear Plant - Specification N2G-877 for Identification of Structures, Systems and Components Covered by the Sequoyah Nuclear Plant Quality Assurance Program.

O Welding: AWS as implemented by TVA General Construction

< Specification G Process Specification for Welding, Heat Treatment, Nondestructive Examination, and Allied Field Fabrication Operations.

O Construction Procedures O SNP Inspection Procedures i

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ENCLOSURE 2 Response to TIR Item Number 11 Identified in August 7, 1986 Letter from NRC TIR Item 11 Provide a description of the background studies or evaluations which were performed to demonstrate that the alternate analysis is equivalent to or more conservative than the rigorous analysis. If it is less conservative, provide the basis for the acceptability of the alternate analysis.

TVA Response Alternate analysis (cookbook) criteria is generated by enveloping the results of several conservative models analyzed by rigorous analysis techniques. Maximum pipe stress and support reactions duc to earthquake loads are typically determined by performing a response s:Jactral analysis of a simple supported one span, two span, and three span model and enveloping the results. Conservative equipment nozzle and anchor reactions are determined by performing a response spectral analysis of a single span fixed-simple support model. Response spectra enveloping several building elevations and several buildings are used. The resulting envelope spectra is modified by extending the peak of the response across the flexible range. This results in all the flexible modes being evaluated at the peak.

The number of supports required on pipe qualified by alternate criteria can typically be reduced by 30 percent or more using a rigorous analysis.

Installed piping qualified by alternate analysis is frequently rigorously analyzed to upgrade documentation. Stress in the piping and pipe support reactions due to seismic loads predicted by rigorous analysis are considerably less (50 to 70 percent less) than those predicted by alternate analysis.

WAM:SFH 8/28/86 0281h i

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ENCLOSURE 3 Submittal of Pages Inadvertently Omitted from August 18, 1986 Letter to NRC on Interim Acceptance Criteria

I. Introductign As e rasult of TVA-NRC m2etings held in Bethesda, Maryland, on July 17-18,1986 to discuss the provisions and use of Civil Engineering

_ Interim Acceptance Criteria as part of the SQN Restart Program, a list of twelve Technical Information Requests (TIRs) was developed. TVA committed at the mcating to expeditiously respond to the requests.

II. Purpose and Scope The purpose of this transmittal is to provice the information NRC requested and expedite the approval of the interim acceptance criteria.

The twelve TIRs are presented and cross-referenced to the pertinent responses.

III. Technical Information Recuests (TIRs) 1.

Provide a description of the analysis procedures being used for cable tray supports interim criteria. Compare these to the original analyses. Include the technical basis for the interim criteria analysis.

Response in Sections IV. B.2, C.2, and D.2.

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Describe how the DBA response spectra on the SCV have been developed. How many cable tray supports are affected by the DBA

.a spectra?

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l Respense in Sec tions IV. B.1, B.3, IV.C.I . , and IV.D.I .

3. Provide description of how anchor bolt load capacities are determined by TVA.

4 Response in Section VI.

4. How will employee concerns be integrated and resolved for the cable tray and piping programs?

, TVA's Employee Concern Program is defined in detail in the Nuclear Performance Plan (revised) for Sequoyah. This plan was transmitted to NRC in a letter dated July 17,1986, from S. A. White to the Honorable Lando W. Zech, Jr.

Employee concerns in the cable tray support area and the small bore piping area have been identified. The subject of several of these employee concerns are being addressed in the cable tray and small bore piping programs. All of the employee concerns involving cable tray supports and small bore piping that affect safety will be addressed and resolved prior to restart of the plant.

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5.

Provida TVA's best estiente of our schtdule for Phase II (af ter plcnt restart) implementation.

Phase II will consist of development of designs, analyses, and technical positions to finalize and document the plant's design basis consistent with FSAR commitments. This process will include additional plant walkdowns which will necessitate staging data gathering with outage schedules.

Ye estimate that, based on presently identified scope-of-effort for the two programs discussed..in this transmittal, that 2 to 3 scheduled outages will be required to obtain data and make the necessary modifications.

Fbase II scope and schedules will be identified prior to restart. .

6.

Provide a list of exceptions to FSAR commitments being taken.

during the interim period (Phase I) and complete justification for interim acceptance criteria.

Responses in Section IV.A. for Cable Tray Supports and Sections V. A. and V.B. for Small Bore Piping and Pipe Supports.

7.

What QA/QC procedures were applied to the original cable tray design and construction?

For the Phase I and Phase II programs?

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Responses in Section VII.

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Mske evciltble to tha NRC rcvitwers copies of all dccumento pertinent to cable tray design and construction as well as cookbook analysis techniques. Material is to be available July 21,1986.

The following material has been provided to NRC reviewers.

TVA drawings 48N1501 through 1506 (supports on SCV)

Walkdown Criteria SMI-0-317-36 TVA drawings for supports in auxiliary and control building CEB Report 80-5 SQN Alternate Criteria for Piping Analysis and Support CEB Report 76-5 WBN Alternate Criteria for Piping Analysis and Support SQN Alternate Analysis Review Program Procedures 9.

Describe the walkdown procedures for data gathering to ensure that the interim cable tray qualification program addresses as-built conditions.

Responses in Section IV. B.4. , C.3. , and D.3.

10.

Provida justification for separating secondary stresses or provide alternate technical approaches for SAM / TAM issues.

4 Responses in Segtion V.A.

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Prcvida cdditionni justification for 2Sy stress level used in faulted primary stresses.

Response in Section V.A.

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