IR 05000327/1986045

Errata to Insp Repts 50-327/86-45 & 50-328/86-45 on 860721-25.Major Areas Inspected:Design Baseline & Verification Program.Page A-8 Inadvertently Omitted from Original Submittal

ML20215N554
Person / Time
Site:	Sequoyah
Issue date:	10/31/1986
From:	Architzel R, Imbro E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:
Shared Package
ML20215N277	List: IR 05000327/1986045 ML20215N276 ML20215N278 ML20215N282
References
50-327-86-45, 50-328-86-45, NUDOCS 8611060324
Download: ML20215N554 (24)

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U.S. NUCLEAR REGULATORY COMMISSION

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OFFICE OF INSPECTION AND ENFORCEMENT Division of Quality Assurance, Vendor, and Technical Training Center Programs Report Nos.: 50-327/86-45, 50-328/86-45 Docket Nos.: 50-327; 50-328 Licensee: Tennessee Valley Authority 6N, 38A Lookout Place 1101 Market S Chattanooga, TN 37402-2801 Facility Name: Sequoyah Nuclear Plant, Units 1 & 2 Inspection At: Knoxville, TN Inspection Conducted: July 21-25, 1986 Inspection Team Members:

Team Leader: R. E. Architzel, Senior Inspection Specialist, IE Mechanical Systems: F. Mollerus, Consultant, Mollerus Engineering In Mechanical Components: A. V. duBouchet, Consultant Civil / Structural: A. Unsal, Consultant, Harstead Engineering Electrical Power: ' V. Athavale, Inspection Specialist, IE Instrumentation &

Control: L. Stanley, Consultant, Zytor In Nuclear Systems: J. M. Leivo, Consultant Operations: P. E.'Harmon, Resident Inspector, SQN Licensing: J. J. Holonich, Project Manager, NRR F. Rinaldi, Structural Engineer, NRR D. Terao, Mechanical Engineer, NRR Ralph E. Architzel Date Team Leader Eugene V. Imbro Date Section Chief Quality Assurance Branch 8611060324 861031 PDR ADOCK 05000328 G PDR

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LIST OF ABBREVIATIONS AEC Atomic Energy Commission ANSI American National Standards Institute APM Anchor Point Movement CFR Code of Federal Regulations CGCS Combustible Gas Control System C/R Commitment / Requirement DBD Design Basis Document DBVP Design Baseline and Verification Program

.DNE Division of Nuclear Engineering EA Engineering Assurance ECN Engineering Change Notice ESF Engineered Safety Features EQ Environmental Qualification FSAR Final Safety Analysis Report GDC 10CFR50, Appendix A, General Design Criteria HVAC Heating, Ventilation and Air Conditioning INP0 -Institute of Nuclear Power Operation .

IEEE Institute of~ Electrical and Electronics Engineers LBB Leak Before Break LOCA Loss of Coolant Accident NEB Nuclear Engineering Branch NMS Neutron Monitoring System NCR Non-Conformance Report NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System P&ID Piping and Instrumentation Diagram RDBD Restart Design Basis Document SCR Significant Condition Report SQEP Sequoyah Engineering Procedure SQN Sequoyah Nuclear Plant TVA Tennessee Valley Authority

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SEQUOYAH NUCLEAR POWER PLANT Design Baseline and Verification Program

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Inspection. Report 50-327/86-45 & 50-328/86-45 July 21-25, 1986 INTRODUCTION AND BACKGROUND The design baseline and verification program (DBVP) was developed by the Division of. Nuclear Engineering (DNE) to resolve design control issues described in several TVA sponsored evaluations and audits and NRC inspection The Sequoyah Nuclear Plant (SQN) Design Baseline and Verification Program will be used by TVA to provide the required level of confidence that the modifica-tions to selected plant systems, implemented since receipt of the operating license, have not resulted in'any violation of the plant's licensing basis. The program'is described in the " Program Plan for the Engineering Assurance Independent Oversight Review for the Sequoyah' Nuclear Plant Design Baseline and Verification Program," dated May 9,.1986 and forwarded to the NRC as an enclosure to Mr. R. L. Gridley's letter dated June 27, 198 . PURPOSE NRC inspection activities related to the TVA's DBVP and associated Engineering Assurance-(EA) independent technical oversight of Sequoyah Nuclear plant are planned to be conducted in several phases:

(1) Inspection of program preparation and initial. implementation (EA Review plans and procedures, DBVP procedures, walkdown results).

(2)' Inspection of program implementation, including' design criteria preparation, Engineering Change Notice (ECN) and system evaluation (3) Inspection of DBVP and EA oversight results and corrective action <

This inspection focused on the development and updating of the design criteri The NRC previously conducted an inspection (Report Nos. 50-327/86-38 and 50-328/86-38) of the DBVP. .This previous inspection focused'on overall j DBVP plan and scope, implementing procedures, and the conduct and results of

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The purpose of this inspection was to (1) review TVA's program for design ~ criteria preparation, (2) review a sample of the revised and newly issued design criteria, and (3) overview the efforts of TVA's Engineering Assurance (EA)

group for independent review of the design criteria. An integral part of the design criteria preparation was the generation and use of a (new) commit-ment /requirelient (C/R) data bas The purpose of the Sequoyah design criteria reconstitution program is to generate revised design criteria documents which address system and general functional design requirements governing the design of structures, systems and components. At TVA, these design criteria documents include current licensing commitments and regulatory requirements, as well as design criteria that are

[ not commitments but TVA self-imposed standards of " good engineering practice."

The revised design criteria will be used by TVA as the basis to review all plant modifications made to those systems or portions of plant systems within

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the scope of the DBVP since operating license issuance to provide assurance that SQN is in conformance with its licensing basis.

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.As a part of the Design Baseline and Verification Program, TVA is preparing a i

Design Basis Document (DBD) and a Restart Design Basis Document (RDBD). The DBD defines, establishes, and maintains the design requirements for the Sequoyah Nuclear Plant. Although a design basis currently exists for the i Sequoyah Nuclear. Plant,- the design basis documents were not always readily retrievable in a verified form. Thus, TVA identified a need for a verified, i controlled design basis document to be maintained throughout plant life. The

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DBD is intended to be used to evaluate and control design changes, to respond i _to abnormal operations and events, to evaluate limiting conditions for i operation, to perform safety reviews, to assess conditions adverse to quality, i to assess operating experience reports, and.to provide an interface with

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outside organizations. The.RDBD will be the initial issue of the DBD and will cover those safety-related systems identified by TVA calculation SQN-OSG7-048 which are required to support hot shutdown and mitigate postulated accidents

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~The DBD contains general design criteria for site, plant, structures, and sys-tems which establish the plant-specific design input requirements. The DBD i also contains certain detailed design criteria, system descriptions, design input drawings, engineering decisions, analysis results, and engineering para-meters for detailed design. The design commitments and requirements to be used as a basis for developing the' design criteria were identified using TVA~

procedure SQEP-18.

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, INSPECTION ACTIVITIES

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The following activities were generally performed by all team member *

Evaluation of applicable DBVP procedures for generation of the commitment / requirement data base and for updating the design basis

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Review of selected design criteria generated or updated as a result of i the DBV *

, Review a sample of the commitment / requirements associated with the l selected design criteria to verify their incorporatio *

l Examination of the results to date of the independent oversight review of C/Rs and design criteria.

t j In addition, the'NRR team members conducted a walkdown at the Sequoyah Nuclear

Plant to physically examine hardware which is the subject of. substantive technical. issues. Observations from the walkdown in the small bore piping and

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HVAC duct and support disciplines are also discussed in this report.

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. SUPNARY OF FINDINGS 4.1 Program for Design Criteria Preparation Team review 'of SQEP-18, " Procedure for Identifying Commitments and Requirements as Source Information for Sequoyah Design Criteria Development," showed that a program has been established to identify the licensing commitments and other design requirements. The team also reviewed SQEP-29, " Procedure for Preparing the Design Basis Document," which addresses the methods to be used to capture the C/Rs and other design input in an upper tier, commitment driven, compila-tion of design and design documentation requirement During a prior inspection of the DBVP (Inspection Report 86-38), the team had performed a preliminary review of the computerized list of licensing commit-ments and design requirements called the " Commitment / Requirement Data Base."

The team found that this list, developed by TVA/Impell, was not independently verified. Although Quality Assurance controls were not applied to the information retrieval process, the Engineering Assurance group has conducted reviews in this area on a sampling basis. The NRC team does not consider that these reviews were done in sufficient technical depth or were of sufficient scope to allow meaningful conclusions to be drawn regarding the completeness of the C/R data base. This data base forms a portion of the basis for the newly issued design criteria document The team discussed future plans the project was considering to verify the accuracy of the C/R data base. Consideration was being given to independently verifying incorporation of all C/Rs in design output documents, for exampl Prior to restart, TVA' considers that the independent verification provided during preparation of new and revised design criteria provides the required verification of incorporation of C/R The team remains concerned regarding the identification of C/Rs. Therefore, TVA should clearly define their basis for concluding that the C/R data base has accurately captured all licensing commitments, regulatory requirements and other pertinent design information as applicable, e.g. NSSS vendor interface requirements (Inspection Report 86-38, Observation No. 5.4).

4.2 Revised and New Issue Design Criteria The team identified a concern regarding the. incorporation of proprietary information in the design criteria due to a lack of availability (within TVA) of proprietary source documents containing C/Rs (Observation No. 3.4)

The team identified several cases where C/Rs applicable to selected design criteria were not captured (Observation Nos. 4.4, 5.5, 6.6, and 6.7).

Environmental Qualification requirements were found to be incorporated in an inconsistent manner among various design criteria (Observation Nos 4.5 and 6.5). The team also identified missing requirements in the design criteria for the neutron monitoring system, apparently at the interface between the NSSS and TV These findings, collectively, raise a concern regarding the compre-hensiveness of the design criteri Several technical observations were identified relating to the coordination (between design criteria) of overcurrent protection and cable sizing for medium voltage motors (Observation No. 5.6); the collection of varied single failure definitions and commitments into a single document (Observation

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No. 6.8); and implementation of 10 CFR 50 General Design Criteria requirements for containment penetrations (Observation No. 8.2)

Several concerns were identified by team members from the NRC Office of Nuclear Reactor Regulation who are reviewing interim acceptance criteria being applied for SQN piping, pipe supports, and cable tray designs (Observation-Nos. 8.1-8.4). These findings are documented in this report for completeness, but will be resolved independently by NRR duriraj their review of these issue During this inspection, the team reviewed several system functional design criteria, as well.as two recently issued Civil Engineering Branch genera ,

design criteria that supersede Watts Bar design criteria which TVA had used to

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design pipe anchors and pipe supports at Sequoyah Nuclear Power Plan The

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team identified two design requirements specified in the Watts Bar design cri-teria for the design and modification of pipe' supports that TVA did not specify t in the Sequoyah design criteria issued on June 23, 1986. The new design criteria does not require that the local stresses generated in the piping by stiff pipe clamps with large preload be checked and allows pipe axial loading

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of floor or wall sleeves that have not been designed for such pipe loads (Observation No. 3.4). -

The Sequoyah Project has not found any items which they determined necessitated the issuance of a new design criteria or any revision to existing design criteria (prior to restart) in the Civil / Structural discipline during review of the licensing commitments and design requirement Several examples of situations which the team considers should have resulted in such criteria changes were identified (Observation Nos. 7.2-7.3).

4.3 EA Independent Oversight Review Program Plan The team expressed certain reservations with the review plan being implemented

for the instrumentation and control discipline to assess the technical adequacy of design changes. The team's concerns included:

i l (1) establishing an appropriate balance between important technical issues and quality assurance issues as described in the checklists;

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i (2) the representativeness of the four systems selected, and

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(3) the relatively small sample sizes chosen for review of design criteria, engineering change notices, and field change notices.

, The instrumentation and control discipline' specific action plan for EA review of the C/R data base states that approximately ten representative commitments /

requirements would be. selected from a variety of so'urce documents including letters and memorandums. These choices will then be reviewed for their inclu-sion in,the Sequoyah commitments / requirements-(C/R) data base. Review attributes described in the implementing checklist include the identification of source

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documents, training of. personnel for this activity, preparation of the C/R data sheet used for input to the Sequoyah commitment / requirement data bask, and dis-tribution of the data base output to involved user While a sample size of ten may be sufficient- to validate .the processing of -

identified commitments and requirements through the Sequoyah C/R data base, the

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, proposed action plan is unlikely to be sufficiently comprehensive to validate the method used to initially identify individual commitments and requirements from the designated source documents. Since the overall effectiveness of the C/R data base is dependent upon the accuracy of the determination of specific commitments and requirements, greater emphasis appears necessary to validate the method and procedure used for identification of commitments and requirement Although no specific observations were identified, team members reviewing'in other disciplines also noted a lack of technical depth and breadth in the samples selected for oversight by EA. Limited oversight products were available in the mechanical systems and nuclear systems areas. In the electric power area, the team found that the EA oversight was thorough and identified findings (Action Items) similar to those identified by the tea In the Civil / Structural area the team found that the plan of action and the attributes shown in the EA oversight review plan indicate that an adequate plan has been established to review the Sequoyah project wor . SPECIFIC COMMENTS Specific comments of individual NRC discipline inspectors are categorized as observation The observations and a description of the activities performed by each discipline of the NRC team are provided in Attachment A of this repor TVA actions relating to individual observations will be reviewed by the NRC during future inspections. These observations elaborate on the general comments stated in this report and in some cases provide additional comments not considered to be of a general natur . MEETING SUMMARIES - REFERENCES A summary of the meetings held relating to the DBVP inspection and a list of references are provided in Attachment B.

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Atttchment A - Insp;ctien Activities and Observations

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NOTE: The observation numbers used in this report are a continuation of the

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numbers used for the previous DBVP inspection. (Report Nos. 50-327/86-38 &

50-328/86-38). The references are listed in Attachment .0 OPERATIONS In the operations area the team examined selected design criteria, EA action items from the previous inspection, incorporation of various commitments /re-quirements, and coordinated with other team disciplines in the conduct of the inspection. No observations were identified in the operations are . MECHANICAL SYSTEMS In the mechanical systems discipline, the team reviewed the following' design criteria generated for the baseline restart effor Responsible Design Criteria N Title TVA Branch SQN-DC-V-3.1.1- Steam Generator Blowdown System MEB SQN-DC-V-4. Main Steam System .MEB

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SQN-DC-V- Feedwater System MEB SQN-DC-V- Essential Raw Cooling Water System MEB SQN-DC-V-13. Auxiliary Buildir.g Ventilation & Cooling MEB SQN-DC-V-13. Auxiliary Feedwater Syste MEB

SQN-DC-V-13. Component Cooling Water MEB SQN-DC-V-1 Auxiliary Control Air System MEB SQN-DC-V-2 ' Residual Heat Removal System MEB The team reviewed the Engineering Assurance Oversight Review Plan - Mechanical (Reference 9). This plan is. organized by the following activitie *

Walkdowns

Licensing Commitments Design Basis Change-Control Board

Evaluation of Change Documents

Comparison of Design Documents with Walkdown Results

System Evaluations

Modifications to Control Room Drawings Unreviewed Safety Question Determination (USQD)

Each activity description includes a plan of action, description of sample size, and an attributes check list. The latter is basically a check list to evaluate the acceptability of the activity being monitore One observation was identified concerning TVA's access to and incorporation of proprietary information in the design criteria (Observation No. 2.3).

Observation No. 2.3 - Status of NSSS Vendor Proprietary Information Several of the commitment / requirements (C/Rs) listed in the analysis report for the component cooling water system are NSSS vendor reports and memoranda that A-1

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are considered by the.NSSS vendor to be proprietary and unavailable to outside

, organizations. The NSSS vendor also stated that the C/Rs contained in proprietary memoranda are also contained in formal documents that have been made available to TVA. The inspection team was informed that the formal documents are being used as a source of C/Rs for the design criteria; the NSSS vendor proprietary correspondence is not. TVA's basis of the acceptability of this approach was the NSSS vendor's statement that all C/Rs are contained in the formal documents. The titles of the proprietary correspondence will be kept in the C/R tracking system until restart. At that time, TVA Branch Chiefs will evaluate all remaining C/Rs not addressed in the design criteri The inspection team will continue to follow developments in this NSSS vendor interface area and TVA's efforts to assure that all NSSS vendor C/Rs are incorporated in the design basis document . MECHANICAL COMPONENTS In the mechanical components discipline, the team reviewed Sequoyah Design Criteria SQN-DC-V-2. 14, Piping System Anchors installed in Category I Structures, which was issued June 30, 1986 and supersedes Watts Bar design Criteria WB-DC-40.31.15 for new pipe anchor designs and modifications to existing pipe anchor design The team also reviewed Sequoyah Design Criteria SQN-DC-V-24.1, Location and Design of Piping Supports and Supplemental Steel in Category I Structures, which was issued June 23, 1986 and supersedes Watts Bar Design Criteria WB-DC-40.31.9 for new pipe support designs and modifications to existing pipe support designs. Observation No. 3.4 was identified relating to elimination of requirements from a design criteria documen Observation No. 3.4 - Pipe Support Design Criteria Sequoyah Design Criteria SQN-DC-V- 24.1, which TVA issued on June 23, 1986 for the design of new pipe supports and modifications to existing pipe support designs, supersedes Watts Bar Design Criteria WB-DC-40.31.9, but does not carry forward the following two design requirements which are contained in the Watts Bar design criteria documen (1) Section 8.2.12 of the Watts Bar Design Criteria, Stiff Pipe Clamps, which notes that stiff pipe clamps with extremely large pre-load may induce sig-nificant localized piping stresses, and requires that piping stresses be investigated for stiff clamps identified in NRC Information Notice 83-8 (2) Section 8.3.1.2 of the Watts Bar Design Criteria, which prohibits pipe

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axial loading of a floor or wall sleeve unless the sleeve has been designed for a pipe loa This issue needs to be further examined by TV . NUCLEAR SYSTEMS In the nuclear systems area the team reviewed TVA's Engineering Assurance (EA) (

oversight work products. At the time of the inspection, the Nuclear Engineer- I ing Branch (NEB) EA products available for NRC review in this area were:

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, . Attachment A - Insp;ction Activities and Observations (1) A review of the design criteria for the combustible gas control system (Reference 4) in accordance with SQEP-29, " Procedure for Preparing the Design Basis Document for Sequoyah Nuclear Plant."

(2) A review of commitment / requirement (C/R) identification for the auxiliary feedwater and essential raw cooling water systems in accordance with SQEP-18, " Procedure for Identifying Commitments and Requirements as Source-Information for Sequoyah Design Criteria Development."

The team also reviewed a sample of DBVP Project generated design criteri At the time of the inspection, approximately half of the NEB design criteria had been issued; none of- the NSSS design criteria had been issue The team reviewed the design criteria issued for the combustible gas control system (CGCS), containment isolation, single failure, and remote shutdown criteria from locations outside control room (Refs. 4 through 7, respectively). In addition, the team reviewed the design criteria for the neutron monitoring system that had recently been transmitted by the NSSS vendor, but had not yet been issued by TVA~.

Three observations were identified relating to incorporation of commitments, consistency of the design criteria, and completeness of design criteria (Observations Nos. 4.4 - 4.6).

Observation No. 4.4 - Spray Shield Commitment / Requirement for Certain Hydrogen Igniters in Upper Compartment A spot check of CGCS C/Rs identified by TVA for commitment changes (and apparently excluded from the analysis report used in preparing design criteria)

indicated that TVA's C/R to provide enlarged spray shields for upper compart-ment igniters (Reference 8) has not been included in the CGCS design criteria (Reference 4). This missing requirement does not appear to have been identified during the EA review of the design criteria. The team recognizes that the igniters are provided for mitigation of accidents beyond the scope of FSAR Chapter 15 events; however, since TVA appears to take credit for the enlarged spray shields in assuring functionality of the igniters following a degraded core loss of coolant accident, the team believes the requirement for enlarged shields should be included in the CGCS/ hydrogen mitigation system design. criteri Observation No. 4.5 - Clear and Consistent EQ Requirements in Design Criteria ;

The team observed that some of the environmental qualification (EQ) require-ments as stated in the few design criteria sampled appear to be somewhat i inconsistent and might be misinterpreted by a user of the criteria who is not particularly knowledgeable of EQ requirements and procedures. For example, the CGCS design criteria do not clearly state what EQ requirements apply to existing equipmen Observation No. 4.6 - Missing Criter.ia for Neutron Monitoring System While the Neutron. Monitoring System (NMS) design criteria had not yet been issued by TVA, the team reviewed the design criteria provided recently by the NSSS vendor (Reference 3). Based on the NMS boundary definition provided in References 1 and 2, the design criteria appear to be incomplete in the following areas:

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l Attachment A - Inspection ActivitiGs and Observations l

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(1) No requirements are stated or referenced for providing indication of

, source range flux outside the control room as required to accomplish safe shutdown from outside the control room; for example, these requirements are not presented in paragraphs 3.5.1.n or Tables 3.1-1, 3.1-4 of the documen (2) No requirements are stated or referenced for electrical penetrations that are required for NHS functionality. An indirect reference to IEEE-317 provided in Section 3.5.2.8 of the document is not considered sufficient to cover the special penetration requirements for these high impedance i instrument <

This observation also reflects the concern expressed in Observation No. (Inspection Report 86-38) that system functional boundaries should be expli-citly defined for this system. For example, it appears that the NSSS vendor has limited the scope of the NMS design criteria to the system and equipment in its scope of supply, rather than addressing the functionality of the integrated ~

system for safe restar TVA should give sufficient attention to the NSSS/Bal-ance-of-Plant interfaces, as well as to overall system functionality, so that no significant items are inadvertently omitted from the review scop . ELECTRIC POWER In the electric power discipline, the team reviewed portions of design criteria SQN-DC-V-11.4.1 (Revision 2), for auxiliary power; SQN-DC-V-11.3 for cable applications; and TVA's DBVP procedures SQEP-16 (Revision 0), SQEP-29 (Revision 1), and SQEP-18 (Revision 1). The team also reviewed the action items identified during EA's independent oversight review of design criteria SQN-DC-V.11.2, SQN-DC-V.11.6, SQN-DC-V-11.8 and DBVP procedure SQEP-16 (Revision 0). The team noted that of the eight design criteria documents required in this discipline prior to restart, the EA oversight team had completed reviews for six desigri criteria documents. This review resulted in the identification of 15 Action Items, similar in nature to the observations identified by the tea Two observations were identified relating to commitment / requirement incorpora-tion and cable sizing (Observation Nos. 5.5 and 5.6).

Observation No. 5.5 - Commitment / Requirement Inclusion in Design Criteria for the Auxiliary Power System The team reviewed portions of Design Criteria SQN-DC-V-11.4.1 (Revision 2), and noted the following weak area (1) Many commitment / requirements have been identified on the commitment /

requirement evaluation sheet but were not captured in the design criteri Examples include the followin SQN EEB DRW 1072 - TVA's commitment pursuant to IE Bulletin 79-25, which prohibits the use of Westinghouse type BFD and NBFD relay SQN EEB DRW 1073 - TVA's commitment pursuant to IE Bulletin 79-11, which prohibits the use of Westinghouse type DB-50 and DB-75 circuit breakers-

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f SQN EEB DRW 1077 - This C/R requires mandatory fusing of control circuit .

SQN EEB DRW.1078 - This C/R deals with problems of spurious operation of diesel generator breaker differential relayin (2) Many commitment / requirements which are applicable to the auxiliary power system were not shown on the commitment / requirement evaluation sheet for this system. Examples includes the followin SQN EEB DRW 1092 - This C/R addres'ses the increased fault interrupting duty of 6.9 kv breakers for 3 phase faults during periods when the turbine generator is operated at higher voltages (25.2 kv) to meet the demands of~

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SQN EEB DRW 1017, 1018, 1019 - These C/Rs address the topic of class 1E

. station battery capacity, and were not included in the auxiliary _ power C/R evaluatio Observation No. 5.6 - Cable Sizing for Overload Currents Section 8.5.3d of Design Criteria SQN-DC-V-11.4.1 (Revision 2), stipulates that

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for medium voltage motors the overload signal will activate an alarm, only. In such situations, since the overload signal does not trip the breaker, the motor feeder cable could carry an overload current from the time of alarm actuation until the breaker is tripped manually. Sizing of the cables shculd address this situatio The team found that design criteria SQN-DC-V-11.3 does not address overload current condition but stipulates the use of 125% of full load current as the basis for sizing the feeder cables. In situations where overload setting is more than 125% of the full load current, the cable sizing may not be adequat This margin of 25% may not be sufficient since the margin may be required for low voltage operation of the motor and by the motor's service factor (capability of supplying greater than rated load). In situations where a motor is required to operate below its normal. rated voltage and/or the motor service factor is more than 1.0, the motor draws more than its normal full load current. The team believes that for cable sizing, the criteria should address evaluation of each i

load on a case by case basis and require sizing the cable accordingly, instead

, of directing cable sizing using a blanket rule of 125% of full load curren . INSTRUMENTATION AND CONTROL The inspection team reviewed TVA procedures SQEP'-18 for identification of commitments and requirements and SQEP-29 for preparation of the design basis document. A detailed review was made of TVA Design Criteria 26.2 for

' environmental qualification, 32.0 for the auxiliary control air system, and 2.16 for single failure criteri Portions of Westinghouse Design Criteria 27.6 for the residual heat removal system, 27.8 for the neutron monitoring system, and 27.9 for the reactor protection system were_ reviewed for their instrumentation and control aspects. In the course of this NRC team review, several system and technical subject output sorts from the TVA data base for commitments and requirements were used as well as a TVA computerized document data base (RIMS) output sorted by the reactor protection system keyword. A number of commitment / requirement record files in the electrical,

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. mechanical, and nuclear disciplines were examined, as well as some source !

documents such as the FSAR, project letters, and plant drawing '

The NRC team also examined instrumentation and control action items identified by the TVA EA oversight team during their review of TVA procedures SQEP-18 and-29 and during their review of the steam generator blowdown and accident i monitoring design criteria. The EA team indicated that one or two more design j criteria would be reviewed; however, the team was concerned with the lack of l depth of technical review being demonstrated by the instrumentation and control '

disciplin Four observations were identified in the Instrumentation and Control discipline relating to specifications for replacement parts, inclusion of C/Rs in design criteria, and definition of the single failure criterion (Observation Nos. 6.5-6.8).

Observation No. 6.5 - Replacement Part and Equipment Qualification TVA Design Criteria SQN-DC-V-32.0 for the safety-related auxiliary control air system has inconsistent criteria for the qualification of Class 1E replacement parts and equipment in that section 3.3.1.f specifies the use of IEEE Trial Use Standard 323-1971. For example, a number of containment isolation valve solenoids for this system are located in a harsh environment, and must now conform to IEEE 323-1974 to satisfy the requirements of 10 CFR 50.4 Although replacement equipment can be exempted from these requirements, sound reasons must be provided as described in Regulatory Guide 1.89. Consequently, the design criteria should be changed to specifically describe those conditions where use of the IEEE 323-1971 standard might be permitte Observation No. 6.6 - Auxiliary Control Air System Design Criteria A comparison was made of the design basis commitments and requirements data base output with the design criteria document for the auxiliary control air system (SQN-DC-V-32.0). In three instances, the team noted that applicable commitments or requirements had not been explicitly converted into design criteria, as follow C/R SQN MEB LWB 1208 stated that an air compressor time delay relay had been replaced with a different relay to achieve a time delay requirement of 3 second This time delay is used to initiate rapid loading of the auxiliary control air compressor shortly after the air header pressure has dropped to 80 psig. The original timer (which had a range of 20 to 200 seconds) did not assure that a minimum 70 psig air header pressure requirement would be maintained. The design criteria document did not reflect this particular time delay requiremen C/R SQN NED FAK 1229 identified a licensee event report (LcR) that contained a statement that TVA would upgrade the air supply to safety-related for containment radiation monitoring containment isolation valves. The design criteria does identify various valves connected to the auxiliary control air system, but does not list the radiation monitor isolaticn valves. The team found that upgrading the air supply to containsent radiation monitor containment isolation valves was not addressed during revision of the design criteria documen A-6

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Attachment A - Inspection Activitics and Obssrvatiens

C/R SQN MEB LWB 1060 identified physical separation requirements for air

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headers inside containment in response to NUREG 0737. A commitment was made that there was no adverse interaction due to a design basis event within containment that would cause a pressure transient that could fail equipment needed to mitigate the transient. The design criteria document did not identify this commitment or the resultant implementation of physical separation' requirements for this syste Observation No. 6.7 - Oil Free Compressed Air Requirement TVA commitment / requirement SQN WES RMM 1110 identified a Westinghouse technical bulletin for solenoid pilot valves in response to NRC IE Bulletin 80-11. In section 4-9 of the Westinghouse SIP /10-1 manual, a requirement was stated in the Westinghouse document that the compressed air system must be oil fre This requirement was omitted from the design criteria document because a satisfactory definition of " oil free" could not be developed in response to an internal TVA review comment. No technical justification was provided that use of a 5 micron filter would suffice to satisfy the oil free requiremen Observation No. 6.8 - Single Failure Design Criteria The team reviewed the single failure criteria document, SQN-DC-V-2.16, dated 7/14/86, with respect to the commitment / requirement data base outpu Three types of concerns were noted by the team; (1) the collection of single failure criteria from various sources has led to a loss of clarity regarding .

certain constraints for application to both fluid and electric systems; (2) the resolution process used during review of the document prior to issuance has produced some definitions that are'no lenger sufficient, and (3) some specific single failure criteria have not been captured in design documents from the commitment / requirement data base outpu The single failure design criteria document did not address each of the initial assumptions of: (1) a loss of offsite power, (2) a design basis event, and (3)

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the consequential failures that are a direct result of that event prior to the application of an individual single failur The design criteria identified the possibility of passive component failure in electric systems as stated by the definition section in 10 CFR 50 Appendix However, the concept of passive failure has been applied only to mechanical fluid systems where certain failures are not postulated during the short-term period. Industry practice has been that postulated failures in electric systems make no distinction between short-term and long-term periods, and are generally identified as active failures to minimize confusion.

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When an undetectable failure exists that cannot be detected by periodic surveillance tests, IEEE Standard 379-1977, endorsed by Regulatory Guide 1.53, states that the preferred course of action is to redesign the system or the test scheme to make the failure detectable. The preferred course of action outlined in the TVA design criteria was to revise the test scheme to eliminate the undetectable condition, and if this could not be achieved, then a redesign may be considered. The team considered that TVA design criteria is too permissive since it appears to encourage the retention of undetectable failures in the system desig The TVA design criteria's definition of independence is "the freedom from effect of one train of equipment on another train." This definition does not A-7

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Attachment A - Inspxction Activities end Obs::rvations

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address the loss of independence that can result from safety system inter-

, actions with either the control systems or with other non-safety-related ..

equipmen '

The TVA. design criteria definition of train is "an arrangement of system and components connected in an (sic) order to provide a dependable function." This definition is considered too broad by the team since it does not distinguish between non-safety-related equipment nor address the existence of redundant equipment. The definition of train should address Class 1E electric equipment and Safety Class 1, 2 or 3 mechanical equipment, and should also require that a train be independent of other redundant trains and connections to non-safety-related equipment that may cause adverse interaction TVA commitments SQN NED RSM 3252 and 3264 stated that Sequoyah would provide diversity as well as redundancy for residual heat removal isolation valve high pressure interlocks to preclude an inadvertent overpressurization of low pressure portions of this system. Design Criteria SQN-DC-V27.6 for the residual heat removal system was developed for TVA by Westinghous This criteria identified the redundancy requirement, but did not include the TVA diversity commitment described F' FSAR amendment 4 Rather, Westinghouse stated that the TVA diversity ummitment had not yet been incorporated into the residual heat removal system design criteria since further review was require This commitment should have been captured in the TVA single failure design i criteria document or the system design criteria, but was not reflected in either documen TVA commitment SQN NEB RSM 5185 identified single failure requirements for the implementation of Regulatory Guide 1.97. In the 1982 TVA licensing submittal on this topic, post accident monitoring separation divisions needed to assure compliance with the single failure criterion were described. The TVA single failure design criteria document did not contain this accident monitoring i

instrumentation commitmen . CIVIL / STRUCTURAL-In the Civil / Structural discipline, the NRC team reviewed the EA independent oversight activities, as well as various TVA design criteria and the computer listing of the commitments / requirements related to all system The NRC team reviewed the following TVA design criteria:

(1) SQN-DC-V-1.3.4, Design Criteria for Category I Cable Tray Support Systems, Rev. O, 8/20/7 (2) SQN-DC-V-1.1.1, Design Criteria for Reinforced Concrete-Block Walls, Re , 1/28/8 '

(3) DS-C1.7.1, Civil Design Standard, General Anchorage to Concrete, Rev. 3, 11/16/8 Interim criteria changes related to these documents were not reviewed by the team, because the Office of Nuclear Reactor Regulatico is reviewing these criteria. The team's review of the computer listing of the commitments and requirements found that no commitment or requirement was identified by TVA

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which would revise the design criteria for the restart phase of the DBVP j (Observations Nos. 7.2 and 7.3).

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Attachment A - Inspzeticn Activities cnd Observations

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', Observation No. 7.2 - Commitm:nts/Requirtments Related to Drilled in Anchors l J TVA engineers reviewed the computer listing of the commitments / requirement data base related to all systems. In this review they determine whether a

- commitment / requirement should be included in the restart phase of the DBV TVA excluded most of the commitments / requirements which relate to drilled-in anchors and placing supports on block walls. TVA's basis for not revising the design criteria prior to restart was that corrective actions te check adequacy in this area were being independently pursued by a significant condition repor The NRC team believes that these commitments / requirements should be re-evaluated to determine whether they should be included in the design

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criteria for the restart phase because they relate to TVA design standard

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DS-C1.7.1 which was recently revised to cover interim criteri J Observation No. 7.3 - Revision of Design Criteria for Restart Te'am review of commitment no. SQNCEB-CG1170 from an "all systems" commitment /

requirement data base computer listing shows that this item was included in the restart phase and that a revision to design criteria SQN-DC-V-1.1.1 was necessar The commitment addresses attachments to reinforced masonry block walls, which are only to be made with through wall bolts. No revision to the design criteria has been made and there was no schedule to revise this design criteria before restart to incorporate the C/R. TVA should reexamine the s

incorporation of C/Rs prior to restart in the Civil Engineering Branch are . -LICENSING In the licensing area, the team focused on the review of TVA newly developed interim. design criteria and related engineering design evaluation, principally in the Civil Engineering area (Civil / Structural and Mechanical Components disc.iplines in this report). The acceptability of TVA's interim acceptance criteria is still under review by the NRC's Office of Nuclear Reactor Regula-tion. 'One day was spent at the Sequoyah plant site, where a walkdown of cable tray and piping systems in containment was conducte OnJulyh3,1986,theNRRteammembersconductedawalkdownattheSequoyah Nuclear Plant. The team was aware of extensive reanalysis of small bore piping systems currently underway at Sequoyah due to several substantive technical issues identified by TVA in their SCR/NCR restart activities. Accordingly, the purpose of the walkdown was to gain a better understanding of the technical issues by viewing specific examples of those designs where the concerns are evident. The team also observed HVAC support designs and typical duct span lengths. The team walkdown primarily included areas inside reactor containment and in the, auxiliary buildin Four observations were identified in the licensing area relating to anchor point movement loads, incorporation of 10 CFR 50 General Design Criteria (GDC)

_ requirements in TVA's design criteria and the results of the site walkdown of

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cable tray, piping, and HVAC systems (Observations Nos. 8.1, 8.2, 8.3 and 8.4).

, All observations stated below constitute open licensing issues that will be

'\ independently resolved by NRR and the license Consequently, these are considered closed for the purpose of this inspectio _

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Atttchment A - InspIction Activiti7s and Observations

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Observation No. 8.1 - Anchor Point'Hovement Loads y

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The team found that anchor point nov'ement (APM) loads associated with a design i basis accident (double-ended guillotine break in the reactor coolant loop) were i

t included in the Watts Bar Design Criteria WB-DC-40.31.9 (previously used fo the design of Sequoyah. supports) but were not included in the new Sequoyah Design C.riteria SQN-DC-V-24.1. 'The licensee's response to the team's observation was that DBA anchor point movement loads were not a licensing issue at Sequoyah. Primary system movements at branch lines resulting from the DBA l loading were found to be less than 1/4 inch. Because the APM results in a secondary stress in the attached branch lines in a plant faulted condition, the

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licensee concluded that the ASME Code does not require its evaluatio !

! According to the licensee, this was a Westinghouse-NRC negotiated positio Furthermore, according to the licensee, this load case does not occun under the Leak-Before-Break (LBB) position _recently approved by the NRC. ' ,1 Although secondary stresses in the piping due to plant faulted conditions ( SSE loading) are generally neglected because they are not expected to cause gross structural failure due to localtyielding and minor distortions in the pipe, the licensee should provide the documentation of the bacis for concluding that the anchor point movements associated with a DBA event are sufficiently small to produce secondary, self-limiting type stresses. Under the Leak-Before-Break position recently established by the NRC, the issue of anchor point movement associated with.a DBM becomes moot. However, the licensee must i ensure that the associated requirements established in the LBB position are l

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adequately implemented in its plant. The NRC office of Nuclear Reactor Regulation sent a letter to TVA (S.A. White from J. Youngblood, dated September 29, 1986), requesting additional information relating to the treatment of anchor point movement loads for Sequoya t Observation No. 8.2.- Conformance to GDC for Containment Isolation s

Appendix A to Design Criteria SQN-DC-V-2-15, " Containment Isolation System,"

states that the Sequoyah design was recently questioned by an NRC inspector who cited the utility for noncompliance with the General Design Criteria (GDC).

Because the basis for the original design criteria for containment isolation-was AEC criterion 53, TVA referenced this in SQN-DC-V-2.15 and intends to remove the GDC reference in the Final Safety Analysis Report and replace it with a reference.to AEC Criterion 53. Since the Sequoyah operating license application date (December 9, 1973) falls more than six months after the effective date of the GDC (May 21, 1971), the plant must meet the GDC; therefore, the licensee cannot change the references, but rather must demon-strate compliance with the GD \ 4 Observation No. 8.3 - Cable Tray System r The following concerns were identified by the team relating to cable tray support system

(1) Review of cross sections of cable trays idOtified cable masses outside the trays. Consideration should be;given to securing the cables and to evaluating resulting eccentricities! An example:of this condition was found in the plant at Elevation'714and Coordinates A12 (2) A few locations in the cable tray systems appeared to contain large unsup-ported spans. These locations were identified to technical personnel

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,Atttchme'nt A -. InspIction Activitics and Observations

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.J involved in the evaluation of these systems accompanying the tea This situation was noted at elevation 714 and coordinates A1Q and A2Q, at elevation 734 at unspecified coordinates, and at elevation 685 in the auxiliary relay room.

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(3) Cables were observed dropping from conduits and cable trays to other cable trays in large bundles without significant support along the vertical drop-span. This was observed particularly in the cable spreading room, but also at other locations. TVA indicated this situation was being evaluate Results of the analytical and testing. programs will be reviewed by the-NR (4) Other concerns identified during the site visit were identified by TVA personnel as currently being evaluated. These concerns include the following:

Supports located near free edges and connections near edges of support plates (Action Item 86-07).

Punch-shear problems between large vertical support tubes and lateral support tubin Correction of weld connection.s between main and secondary structural member (5) Various team discussions with TVA technical staff did not clarify how TVA would assure that all employee concerns applicable to the interim criteria would be evaluated.

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The NRC Office of Nuclear Reactor Regulation is evaluating cable tray support systems at Sequoyah. A request for additional information was sent to TVA (Mr. S.A. White from J. Youngblood, dated August 28, 1986) on this subjec Observation No. 8.4 - Piping and HVAC Systems For small bore piping inside containment, the NRC observed several large, extended motor-operators on valves which were unrestrained. These large

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motor-operated valves were found to be extremely flexible and sensitive to dynamic excitations. This generic issue was previously identified by TVA in a recent SCR (SCRSQNCEB8614). TVA found that the effect of torsion due to seismic acceleration of large eccentric masses (e.g., valve operators) was at *

Ltimes neglected. The safety implications were that excessive pipe stresses and pipe support loads could result. Additionally, excessive displacements could potentially impair the operability of the valve and cause damage to adjacent, sensitive equipment. In 1979, a program for review of unsupported valve ~

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operators resulted in the correction of several deficiencies in this are However, as documented in the above noted SCR, some problems may still exis Motor-operated and pneumatic valves will be reviewed in the alternate program and deficiencies will be evaluated and correcte The second issue observed by the NRC relates to non-seismic piping inside

- containmen The team observed a large portion of a small bore piping system

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which was not supported in the horizontal direction. The licensee believed the line to be non-seismic and, thus, should be addressed in the seismic (II/I)

interaction program. There currently appear to be at least two programs in place at.TVA which may address this potential issue.

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. Attichment A - InspIcticn Activitics and Observations

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Alternately analyzed Category I (L) type non-ANS ~ safety cl. ass line (those performing a secondary safety function) are being evaluated by the TV mechanical discipline through a contract with EQE Inc., which performs walkdowns to assess the seismic integrity of the piping syste Secondly, the staff is aware of a Sequoyah probabilistic seismicity calculation (Analysis No. SQNNAL5-003) and an evaluation by Stevenson &

Associates which calculated the probability of a seismic event inducing a failure in a non-seismically designed pip This observation was identified to document and track resolution of these issue The NRC Office of Nuclear Reactor Regulation is~ evaluating TVA's alternate analysis programs and interim acceptance criteria and will follow

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resolution of these issues.

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, Attachment B'- Me: tings and ReferInc:s

. Meetings Inspection activities were conducted at the DNE offices in Knoxville, Tennesse Entrance and exit meetings were held to discuss the inspection plans and findings, respectively. The following describes the general purpose of these meetings. Table B.1 is provided as a matrix of meeting attendance and principal persons contacted. Other licensee personnel were also contacte Meeting 1: .On July 21, 1986, an entrance meeting was held at the DNE offices in Knoxville. The NRC explained the plans for the assessment of TVA's program for design criteria preparation and the associated Engineering Assurance oversigh Meeting 2: On July 25, 1986 an exit meeting was held at the DNE offices in Knoxville. The scope and findings of the inspection were discussed. The team members presented the more significant findings within each disciplin Table A.1 MEETINGS Name Organization Title Meeting Attended 1 2 REArchitzel USNRC-IE Team Leader x x SVAthavale USNRC-IE NRC-Electric Power x x .

PEHarmon USNRC-RII Resident Insp.,SQN x ADuBouchet NRC-Consultant NRC-Mech. Components x x FJMollerus NRC-Consultant NRC-Mech. Systems x x AIUnsal NRC-Consultant NRC-Civil / Structural x x JMLeivo NRC-Consultant NRC-Nuclear System x x LStanley NRC-Consultant NRC-Instr./ Controls x x WCDrotleff TVA-DNE Dir. DNE x x JEHuston TVA-DNQA Dep. Dir. Nuc. QA JFWeinhold TVA-DNE EA Manager x APCappozzi TVA/S&W Consultant - EA x x MPBerardi TVA-EA EA Oversight Team Lead x x AWLatti TVA-DNE Manager DBVP x x

'RPSvarney TVA-EA Civil / Structural Eng x x APagano TVA-EEB EEB Asst. BC x JPLittle TVA-MEB Supervisor x JPDurnhan Impell Consultant x JFCox TVA-DNE Ast. PE SQEP-K- x x MJScruggs TVA-DNE Elec. Engineer SQEP-K x x CFBowman TVA-DNE DPB Mg x JJSas TVA-DNE Dep. Director DNE x JARaulston TVA-DNE Chief Nuc. Eng x FAKoontz TVA-NEB Gr. Head T/H & Plant Su CWParker TVA-NEB Nuc. Eng x JJWilder TVA-NEB Nuc. Eng x BHall TV Licensing - Sequoyah x x GRReed TVA-DNE Elec. Engr. - DBTF x RL0lberding TVA-DNE Mech. Eng x RCWilliams TVA Reg. Re x DLWilliams TVA Nuc. Eng x B-1

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Attachment B - Me2 tings and Refarences

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-Table A.1 MEETINGS (cont.)

Name Organization Title Meeting Attended 1 2 JWilliams TVA Staff Specialist x FRinaldi NRC-NRR ' Structural Eng x x DTerao NRC-NRR Mechanical Eng x x JHolonich NRC-NRR Project Mg x REFERENCES TVA Calculation SQN-0S67-048, Revision 1 dated 5/15/86,

" Identification of Systems Required for Restart." " Design Baseline and Verification Program, Sequoyah Nuclear Plant,"

Revision 0 dated 5/1/8 . TVA Design Criteria SQN-DC-V-2).8 and " Status of C/R Review" for

" Neutron Monitoring System," transmitted by Westinghouse letter dated 7/16/86 to TVA for information and use; not yet issued by TV . TVA Design Criteria SQN-DC-V-26.1 Revision 0 dated 7/11/86,

" Combustible Gas Control System." TVA Design Criteria SQN-DC-V-2.15 Revision 0, " Containment Isolation System." TVA Design Criteria.SQN-DC-2.16 Revision 0, " Single Failure." TVA Design Criteria.SQN-DC-2.17 Revision 0, " Remote Shutdown Criteria from Locations Outside the Main Control Room." TVA letter 11/2/84, Mills to NRR, topic: Hydrogen Igniter Spray Shield Desig . Review Plan No. 4100R2, Tennessee Valley Authority, Sequoyah Nuclear Plant, Engineering Assurance Oversight Review Plan, Mechanical, Jul/ 23, 198 .

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, NRC QA REVIEW ACTIVITIES

TECHNICAL REVIEW, OVERSIGHT, AND AUDIT / AUDIT OBSERVATIONS

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Techincal oversight and surveillance by line organization of contractors and subtier organizations -- adequate staffing and technical competence

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Auditing by QA or other independent organization with heavy technical orientation (" vertical slice" techniques)

as opposed to focusing mainly on QA process / procedures

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Technical capability of reviewers / auditors -- capable of performing work being reviewed. QA not staffed just by "QA" specialists

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Strong corrective action program

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" Vertical slice"/end product assessment techniques for deter-mining quality program (both doers and reviews) effectiveness TECHNICAL PLANS AND PROCEDURES

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Work performed, and technical procedures developed, consistent with approved technical program plans DOCUMENTATION

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Need for auditable trail

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Use of documented procedures in technical work Enclosure

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Mr. C. C. Mason -4-

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g Distribution (w/ encl)

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'DQAVT Reading QAB Reading RArchitzel, IE LSpessard, IE EVImbro, IE

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BKGrimes, IE JMTaylor, IE RWStarostecki, IE HRDenton, NRR GZech, RII KBarr, RII BBHayes, OI BDebbs, RII , ,

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Inspection Team (8)

JHolonich, NRR !

Resident Inspector NSIC Regional Administrator, RII

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CRStahle, NRR l TMNovak, NRR L NTIS ELD OGC L  % $ ,4 IE:DQAVT:QAB IE:DQAVT:QAB IE:DQAVT:'QAB:C IE:D :DD REArchitzel EVImbro HMiller H ler 10f//86 10/ //86 10/g/86 0/ /86 Qp .D

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D cs Please discard previous edition of this memorandum. Page A-8 was inadvertently omitte .

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