ML20085L365
| ML20085L365 | |
| Person / Time | |
|---|---|
| Site: | Marble Hill |
| Issue date: | 09/30/1983 |
| From: | TORREY PINES TECHNOLOGY |
| To: | |
| Shared Package | |
| ML20085L363 | List: |
| References | |
| GA-C17258, GA-C17258-V02, GA-C17258-V2, NUDOCS 8310240179 | |
| Download: ML20085L365 (129) | |
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L GA-C17258 -
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{ INDEPENDENT CONSTRUCTION REVIEW OF
, MARBLE HILL NUCLEAR GENERATING STATION UNITS 1 AND 2 l
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VOLUME 2 l PROGRAM RESULTS l
l Prepared for l Public Service Indiana l
l GA PROJECT 2485 l SEPTEMBER 1983 l
J otocy A Di.ison o' G A Technologies Inc. unusuuuuum R8A!!8aPo!880s!6 A PDR
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i FOREWORD This final report is presented in three volumes:
Volume 1. Executive Summary Volume 2. Program Results Volumu 3. Potential Finding Reports
( Volume 1, Executive Summary, is a complete overview of the program, the work performed, and the major conclusions drawn.
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Volume 2, Program Results, provides a thorcugh discussion of the entire program. It describes the program organization, the actual work performed, the questions raised during the review, the resolution of these questions, the final conclusions associated with each part of the program, and the overall program conclusions.
Volume 3, Potential Finding Reports, is a compilation of all of the Potential Finding Reports initiated during the review, the Corrective
( Action Plans for the Findings, and a review of those Corrective Action Plans. Volume 3 does not include program discussions, a description of the work, or any conclusions.
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LIST OF ACRONYMS AFS Auxiliary Feedwater System ,
ANI Authorized Nuclear Inspector ANSI American National Standards Institute AQAM American Society of Mechanical Engineers Quality Assurance Manual f
ASME American Society of Mechanical Engineers ASTM American Society for Testing Materials
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BOP Balance of Plant (system)
CAP Corrective Action Plan CBI Chicago Bridge and Iron Company CCWS Component Cooling Water System CFR Code of Federal Regulations CMTR Certified Material Test Report DDIS Design Document Information System DRN Document Review Notice ECCS Emergency Core Cooling System ECN Engineering Change Notice ERA Engineering Release Authorization FCR Field Change Request FSAR Final Safety Analysis Report GA GA Technologies Inc. (formerly General Atomic Company)
Inryco Inland Ryerson Company
( LOCA loss of cooling accident MHNGS Marble Hill Nuclear Generating Station MMIS Materials Management Information System
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NCR Nonconformance Report NDE nondestructive examination NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System
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LIST OF ACRONYMS (ContinucO NUREG Nuclear Regulation ODO Original Design Organization PFR Potential Finding Report PI Project Instructions P&I piping and instrumentation P&ID piping and instrumentation diagram PMP Project 1:anagement Procedures P.O. Purchase Order Project Quality Assurance Manual I PQAM PSAR Preliminary Safety Analysis Report PSI Public Service Indiana QA Quality Assurance QAP Quality Assurance Procedure QAPD Quality Assurance Program Document QC Quality Control QCP Quality Control Procedure QE Quality Engineering RCS Reactor Coolant System RHR Residual Heat Removal System S&L Sargent & Lundy Engineers SMI Storage and Maintenance Instruction (number)
TPT Torrey Pines Technology I
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- CONTENTS FOREWORD . ............................ iii LIST OF ACRONYMS . . .... .. ................. v
- 1. INTRODUCTION . . .... .. ................. 1-1 1.1. Obj ective ...... ................. 1-1 1.2. Program StructuZe .. ................. 1-2 1.3. TPT Qualifications and Independence . . . . . . . . . . 1-3
( 1.4. Evaluation Process . . ............. . . . . 1-5 1.5. Processing Potential Finding Reports . . . . . . . . . . 1-6
- 2. EVALUATION OF QA ORGANIZATION AND MANAGEMENT POLICIES TOWARD
( QUALITY ASSURANCE, TASK A . ............. . .. . 2-1 2.1. Objective and Scope .. ................ 2-1 2.2. PSI Personnel and Documents . . . . . . . . . . . .. . 2-2 2.3. Organizational Level and Status of the PSI QA Department . .. . .. ................. 2-2 2.4. Authority and Responsibility of QA Personnel . . . . . . 2-5 2.5. QA Department Access to Upper Management . . . . . . . . 2-6 42.6. QA Department Involvement in Project Activities . .. . 2-8 2.7. Management Involvement in QA Activities . . . . . . . . 2-10 2.8. Conclusions, Task A . ................. 2-12
- 3. CONSTRUCTION DESIGN CONTROL, TASK B . . . . . . . . . . .. . 3-1 3.1.
( Task Objective and Scope . . . . . . . .. . . . . . . . 3-1 3.2. Design Change Control Procelure Review . . . . . . . . . 3-1 3.3. Procedure Implementation Review . . . . . . . . . . . . 3-5
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3.4. Conclusions, Task B . ............. . . . . 3-5
- 4. PHYSICAL VERIFICATION-WALKDOWN, TASK C . . .. . . . . . . . . 4-1 4.1. Task Objective and Scope . . . . . . .. . . . . . . . . 4-1 4.2. Feature Selection . . ................. 4-2 4.3. Walkdown Procedures . ...... . . . . . . . . . . . 4-3 4.4. Physical Verification Performance . . . . . . . . . . . 4-4
( 4.4.1. Verification of Mechanical Systems . . . . . . . 4-5
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4.4.2. Verification of Civil Structures . . . . . . . . 4-10 4.4.3. QA Support for Walkdown . . . . . . . . . . . . 4-13 4.5. Conclusions, Task C . . . . . . . . . . . . . . . . . . 4-14
- 5. TESTING, TASK D . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. ASME Piping Weld Inspection, Subtask D1 . . . . . . . . 5-1 5.1.1. Objective and Scope . . . . . . . . . . . . . . 5-1 5.1.2. Review of Relevant Marble Hill Procedures . . . 5-1 5.1.3. Weld Inspection Review . . . . . . . . . . . . . 5-2 5.1.4. Conclusions, Subtask D1 . . . . . . . . . . . . 5-4 5.2. Concrete Inspection, Subtask D2 . . . . . . . . . . . . 5-5 5.2.1. Objective and Scope . . . . . . . . . . . . . . 5-5 1
5.2.2. Procedure . . . . . . . . . . . . . . . . . . . 5-6 5.2.3. Concrete Strength Tests . . . . . . . . . . . . 5-7 5.2.4. Conclusions, Subtask 02 . . . . . . . . . . . . 5-8 5.3. Overall Conclusions, Task D . . . . . . . . . . . . . . 5-8
- 6. CONSTRUCTION DOCUMENT REVIEW, TASK E . . . . . . . . . . . . . 6-1 6.1. ASME Piping:faterial Certification Review, l Subtask E1 . . . . . . . . . . . . . . . . . . . . . . . 6-1 1 6.1.1. Objective and Scope . . . . . . . . . . . . . . 6-1 6.1.2. Selection of Certified Material Test Reports . . 6-1 6.1.3. Review of Material Cartifications . . . . . . . 6-2 6.1.4. Conclusions, Subtask E1 . . . . . . . . . . . . 6-2 6.2. Concrete Test and Inspection Review, Subtask E2 . . . . 6-3 6.2.1. Objective and Scope . . . . . . . . . . . . . . 6-3 6.2.2. Selection of Structures . . . . . . . . . . . . 6-3 6.2.3. Specifications and Procedures . . . . . . . . . 6-4 6.2.4. Test and Inspection Records Review . . . . . . . 6-S 6.2.5. Conclusions, Subtask E2 . . . . . . . . . . . . 6-7 6.3. Welder Qualification Records Review, Subtask E3 . . . . 6-7 6.3.1. Objective and Scope . . . . . . . . . . . . . . 6-7 6.3.2. Selection of Welder Qualification Records . . . 6-7 6.3.3. Review of Welder Qualification Records . . . . . 6-8 6.3.4. Conclusions, Subtask E3 . . . . . . . . . . . . 6-8 viii
6 6.4. Safety-Related Equipment Maintenance and Storage Review, Subtask E4 . . . . . . . . . . . . . . . . . . . . . . . 6-9 m 6.4.1. Objective and Scope .. . . .. ...... . . 6-9 f
- 6. 4. 2., Review of PSI Procedures . .. . .. . .... . 6-9 6.4.3. Selection of Stored and Installed Equipment . . 6-9 6.4.4. Review and Inspection of Equipment . . ... . . 6-10 6.4.5. Conclusions, Subtask E4 . .. .. . ... . . . 6-10 6.5. Overall Conclusions, Task E .. ... . .. . ... . . 6-10
- 7. OVERALL
SUMMARY
AND CONCLUSIONS ... . ... .. . ... . . 7-1
( 7.1. Summary of Observations and Findings . . .. . .. .. . 7-1 7.2. Summary of Corrective Action Plans . .... . .. . . . 7-3 7.3. Conclusions 7-4
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FIGURES 1-1. Procedure for processing PFRs . . . . .. . .. . ... . . 1-10 4-1. Unit 1 RCS components verified . . .... .. . ... . . 4-31 4-2. Unit 1 AFS components verified . . . ... .. . .. . . . 4-32
( 4-3. Unit 1 CCWS components verified . . . . . . . . . . . . . . 4-33 4-4. Unit 2 CCWS components verified . . . . . . . . . . . . . . 4-34 4-5. Unit 1 RHR components verified .. . . .. .. . .. . .. 4-35 4-6. Unit I reactor containment building areas verified .. . . 4-36 4-7. Unit 2 reactor containment building areas verified . .. . 4-37 4-8. Fuel handling building areas verified . . ... . .. . . . 4-38 4-9. Ultimate heat sink structure areas verified . . . . . . . . 4-39 4-10. Unit 1 and 2 auxiliary building areas verified . ... . . 4-40
[ TABLES 1-1.
( List of procedures for independent construction review of Marble Hill Nuclear Generating Station Units 1 and 2 . . . 1-9 2-1. Personnel interviewed in Task A . . . . . .. . ... . . 2-14 2-2. Documents reviewed in Task A . . . . . . . . . . .. . .. 2-15 3-1. Procedural documents reviewed in Task B . . . . . . . . . . 3-6
( 3-2. Design documents reviewed in Task B . . .. .. . .. . . . 3-7
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l TABLES (Continued) l 4-1. Installed pipes physically verified at level 2 for Units 1 a,nd 2 . . . . . . . . . . . . . . . . . . . ... . 4-15 4-2. Installed pipe supports physically verified at levels 2 and 3 for Units 1 and 2 . . . . . . . . . . . . . . . . . . 4-16 4-3. Summary of mechanical features verified . . . . . . . . . . 4-17 4-4. Reactor Coolant System equipment verified at level 2 for Unit 1 ...... ............. . .... . . 4-18 4-5. Auxiliary Feedwater System equipment verified at level 2 for Units 1 and 2 . . . . .. . . .. . . . . . ...... 4-19 4-6. Component Cooling Water System equipment verified at level 2 f or Unit s 1 and 2 . . . . . . . . . . . . . . . . . 4-20 4-7. Residual Heat Removal System equipment verified at level 2 for Units 1 and 2 . . . . . . . . . . ... . .. .... . 4-22 4-8. Summary of civil / structural verification . ... . ... . 4-23 4-9. Constructors and suppliers involved in civil / structural CCDstruction ..... . ... . .. . . . . .. .... . 4-27 4-10. Types of reinforced concrete and items covered during walkdown at levels 2 and 3 ... .. . . . . ... .. .. 4-28 4-11. Types of structural steel frame and items covered during walkdown at levels 2 and 3 .. . . . . . . ... . .. . . 4-29 4-12. Types of cable tray supports and items covered during walkdown at levels 2 and 3 . . . .. . . . . . ..... . 4-30 5-1. NDE procedures reviewed in Subtask D1 . . . . . . .... . 5-9 5-2. Welds selected for Subtask D1 . . . . . . . . . . . . . . . 5-10 5-3. Welds examined in Subtask D1 . . . . ... . .. .. .. . 5-15 5-4. Welds selected for in process inspection in Subtask D1 . . 5-16 5-5. Summary of concrete tests and inspections in Subtask D2 . . 5-17 6-1. Certified Material Test Reports reviewed in Subtask E1 . . 6-11 I 6-2. Summary of Nonconformance Reports reviewed in Subtask E2 .. ........... . . . . .. . .... 6-13 6-3. Welder qualification list for Subtask E3 . . .... . . . 6-16 6-4. PSI procedures for safety-related equipment maintenance and storage reviewed in Subtask E4 . . . . . .. . .. . . 6-17 6-5. Equipment selected for storage and maintenance records verification in Subtask E4 . . . . . . . . . .. ... . . 6-18 7-1. Potential Finding Report Summary . . . . . . . .. .. . . 7-6 k
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- 1. INTRODUCTION -
Torrey Pines Technology (TPT), a division of GA Technologies Inc.
(CA), was engaged by Public Service Indiana (PSI) to conduct an independent review of the construction of Marble Hill Nuclear Generating Station (MHNGS) Units 1 and 2, including the quality assurance (QA) organization and PSI management policies toward QA. This is the final report for that program, which began in June 1983. At the time of this review, Unit I was
( approximately 50% complete and Unit 2 approximately 25% complete.
1.1. OBJECTIVE The independent construction review addressed the following:
- Adequacy of the QA organization and whether the attitude of PSI management toward QA would promote a quality attitude among those performing and checking the construction so that the high standards expected of the nuclear industry would be maintained.
( Control of required design changes by PSI upon receipt from the architect-engineer through construction of hardware.
- Conformance of construction of the MHNGS to engineering documents.
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- The Marble Hill Nuclear Generating Station is located on the Indiana shore of the Ohio River, approximately 30 miles northeast of Louisville, Kentucky. It is a two-unit plant with Westinghouse 3425-MW(t) reactors and Westinghouse turbine generators with a net output of 1132 MW(e) each. The project is a replicate of Commonwealth Edison's Byron and Braidwood Sta-tions. Sargent & Lundy Engineers is the architect-engineer. PSI is the Constructor.
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I The review was structured to provide a discerning basis for judging the adequacy of MHNGS construction.
1.2. PROGRAM STRUCTURE A seven-element approach to construction review was executed as follows:
Task A. Review of the QA organization and management policies toward QA.
Task B. Review of the system for controlling design changes and their implementation during construction.
Task C. Physical verification of plant hardware.
Task D. Testing and inspection of American Society of Mechanical Engineers piping welds and concrete.
Task E. Review of selected construction documents.
Task F. Processing of Potential Findings.
Task G. Administration and reports.
I A program plan was prepared early in the project to define the specific tasks required for the independent evaluation; all the subsequent work was in accordance with that plan.
The program was structured to verify that the construction process adequately converted design document requirements into completed features and that QA was receiving adequate management attention. This was accom-plished by reviewing the QA organization and management policies toward QA (Task A), reviewing the construction design control system and its imple-mentation (Task B), physically inspecting the hardware of the plant (Tasks C and D), and reviewing selected construction documents (Task E). The selection of plant features to be reviewed gave primary consideration to public safety. Any deviation that was identified during this process that s
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potentially could significantly impact the safety of the plant was documented and reviewed (Task F). Status reports and a final report on the adequacy of the Marble Hill plant construction were prepared and issued (Task G). -
The program reviewed the activities of PSI, Sargent & Lundy Engineers (S&L), Westinghouse Electric Corporation, Gust K. Newberg Construction Company and Gust K. Newberg, Incorporated, Joint Venture (Newberg), and Cherne Contracting Corporation. More than 2,100 documents were reviewed or used in the review, and more than 13,000 checks were performed on the implementation of procedures and the verification of hardware. Over a
( four-month period, approximately 200 manweeks of effort were applied to this program.
1.3. TPT QUALIFICATIONS AND INDEPENDENCE TPT, a division of GA, is qualified to review the construction of the MHNGS for PSI because GA has been actively engaged in the nuclear power industry since 1959 and has a large staff of capable, experienced, and technically trained personnel. In addition, GA operates under the first Nuclear Regulatory Commission (NRC)-approved QA program and has acknowl-edged expertise in QA. This independent construction verification for PSI was conducted under the provisions of this QA program.
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GA is a privately owned center for diversified energy research, development, and engineering. Activities have emphasized the creation of advanced systems of power generation and energy conversion.
GA employs approximately 1900 people, of which 900 are degreed
( professionals, including 485 with advanced degrees. Many of the technical staff are recognized leaders and experts in their fields. The staff is
{ highly experienced in the nuclent field and has an extensive background in water-cooled nuclear power plant work.
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Prior to this construction review of Marble Hill for PSI, TPT conducted other independent reviews, such as the seismic design reviets of I San Onofre Units I and 3 for Southern California Edison, a constructiot.
verification review of Shoreham for Long Island Lighting Company, a QA evaluation and design review of Palo Verde for Arizona Public Service Company, and a system design review of Waterford 3 for Louisiana Power and Light. TPT has also just completed a management review of the Zimmer Nuclear Power Station for Cincinnati Gas & Electric.
GA and all its personnel on this program are completely independent of PSI, the managing and operating agent for participants in the MHNGS pro- l ject. GA has not had significant involvement with PSI and the MHNGS in the last two years.
GA has received less than $1 million in revenues in the last two yaars (less than 1% of its annual revenue) from PSI and its Marble Hill plant contractors. The following companies are involved in the Marble Hill plant:
- 2. Sargent & Lundy Engineers (S&L)
- 3. Westinghouse Electric Corporation
- 4. Gust K. Newberg Construction Company and Gust K. Newberg, Incorporated, Joint Venture (Newberg)
- 5. Cherne Contracting Corporation
- 6. Commonwealth-Lord, Joint Venture l
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The individuals involved in this program were free of substantive conflict of interest. Substantive conflict of interest is defined as follows:
- 1. For key project personnel, any work experience or association in design, construction, and QA with the Marble Hill plant or with
( PSI within the past three years.
- 2. Current activity on any other Marble Hill or PSI work.
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- 3. For personnel, other than key project personnel, with past Marble Hill or PSI work experience or association within the past three years, a level of effort on this project which exceeds a half-( time level for the duration of the project.
- 4. An immediate family member who is employed by PSI.
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- 5. A cumulative ownership and creditor interest in PSI which exceeds
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5% of their gross family annual income.
r 1.4. EVALUATION PROCESS Technical and QA personnel were assigned specific items to review.
The reviewers requested the appropriate documents from the original design
( organizations, such as PSI, S&L, and Westinghouse, and performed the neces-sary reviews and evaluations to satisfy the requirements of the program.
When an apparent deviation was identified by a reviewer, which could conceivably result in a safety hazard, the reviewers filed a Potential r Finding Report (PFR) and carried the report through the PFR processing F
procedure to its final disposition.
r L Throughout the review and the processing of PFRs, emphasis was placed on the independence of the reviewer. A reviewer required no approval to
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l file a PFR, and, once filed, the PFR was required to be processed according to the written procedure.
To guide the reviewer in his work, a set of review procedures was prepared for each task and for the processing of PFRs. Table 1-1 lists these procedures. The procedures established the scope of the work and provided rules for conducting the review. Copies of the appropriate procedures were provided to all personnel who were designated as reviewers. I Meetings were held by the task leaders to discuss these procedures with the reviewers. Also, training sessions were held to familiarize reviewers with the procedures for filing and processing PFRs.
A QA Program Document (QAPD) was prepared to describe the QA require-ments governing work under this project. The QAPD was distributed to all project personnel. The QA program included an internal audit performed by the GA QA division to evaluate compliance with the QAPD, with project pro-cedures, and with the program plan. The audits indicated that compliance I was satisfactory.
I 1.5. PROCESSING POTENTIAL FINDING REPORTS l
The main features of the procedures for processing PFRs are shown in Fig. 1-1. Reviewers for Tasks A, B, C, D, and E initiated a PFR when they encountered an apparent deviation that met the definition of a Potential Finding contained in their review procedures. All PFRs were reviewed by the task leader. All PFRs were transmitted to the original design organi- I zation (ODO) to review for accuracy. Independence of the initiator was maintained giving him the sole right to reject or incorporate comments by either the task leader or the ODO. Permanent records of all PFRs, includ-ing revisions, were maintained, and all comments, whether incorporated or rejected, were documented. l All tables and figures are grouped at the end of each section.
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After review by the task leader and the ODO, the PFR was sent (together with an Impact Assessment for valid PFRs) to the Findings Review Committee for evaluation and classification. The Impact Assessment was an appraisal by the'PFR initiator and the task leader of the seriousness of the Potential Finding.
A PFR was classified as valid if, af ter the above-described review, the initiator and the task leader agreed that it was valid or if either party and the Findings Review Committee believed the PFR was valid. A PFR could be classified as invalid if either of the abcve-identified parties concluded that the Potential Finding was invalid and the Findings Review Committee agreed.
Each PFR was reviewed and evaluated through several steps and ultimately classified as Invalid, an Observation, or a Finding.
4 The review procedure contained criteria which required that a valid Potential Finding be classified as a Finding if any of the following conditions existed:
- 1. The design margins had been reduced to the extent that design allowables were exceeded (after allowance for any inherent con-servatism in the analytical method used) or safety-related design requirements were not met.
- 2. An isolated procedural deviation was of such a nature that a substantial safety hazard could be created.
- 3. Repetitive similar procedural deviations existed which, when taken together, implied that a substantial safety hazard could be created.
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- 4. The nature of the deviation was such that nunerous similar deviations may exist and that this situation could create a substantial safety hazard. Examples of this condition are deficiencies in project or design control procedures, overall specifications and/or requirements, and/or commonly used calculational techniques.
- 5. Repetitive deviations resulting from unsuitable methods or errors in specifying, interpreting, or implementing the requirements exist which did not themselves create a substantial safety hazard but the nature of which, considering the size of the sample exam-ined in the review, suggests that other errors or deviations may exist which could create a substantial safety hazard.
A Potential Finding judged to be " valid" but which did not meet any of the above criteria was classified as an " Observation."
A PFR was classified as Invalid if additional information was provided that eliminated the concern subsequent to the initial issuance of the PFR.
The classification of the Potential Finding was reviewed by the Project Manager to determine if the correct procedures had been followed.
Subsequently, the Observations and Findings were sent to the Senior Vice President, Nuclear Division, of PSI. In the case of Findings, a Corrective Action Plan (CAP) was prepared by PSI which identified the cause of the deviation, steps to be taken to remedy the deviation, steps to be taken to identify similar deviations, and steps to be taken to prevent the recur-rence of the deviation. The CAP was reviewed by TPT to determine if it satisfied the concern expressed in the Finding.
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- TABLE l-1 LIST OF PROCEDURES FOR INDEPENDENT CONSTRUCTION REVIEW OF MARBLE HILL NUCLEAR GENERATING STATION UNITS 1 AND 2 Procedure No. Title QAPD-2485 Quality Assurance Program Document - Independent Construction Review of Marble Hill 2485-PD-01 Review of QA Organization and Management Policy 2485-PD-02 Review of Construction Design Control 2485-PD-03 QA Review cf Walkdown Activities 2485-PD-04 ASME Weld Inspection
- 2485-PD-05 Concrete Testing and Inspection 2485-PD-06 Review of ASME Piping Material Certification 2485-PD-07 Review of Concrete Test and Inspection Records 2485-PD-08 Review of Welder Qualification 2485-PD-09 Review of Safety-Related Equipment Maintenance and Storage
- 2485-PD-10 Processing of Findings 2485-PD-11 Physical Verification of Mechanical Features 2485-PD-12 Physical Verification of Structural Features 2485-PD-13 Document Control Procedures t
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- 2. EVALUATION OF QA ORGANIZATION AND MANAGEMENT POLICIES TOWARD QUALITY ASSURANCE, TASK A 2.1. OBJECTIVE AND SCOPE This task evaluated the organizational status of QA on the Marble Hill
( project and reviewed PSI management policies that affect QA to assess the degree to which those policies help assure an effective QA program.
The scope of this task included the following:
o Evaluate organizational level and status of the PSI QA Department.
e Evaluate job descriptions and procedures relevant to defining authority and responsibility of key QA personnel.
e Evaluate PSI QA Department's access to upper management.
e Evaluate PSI QA Department's involvement in project
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e Evaluate management's involvement in QA activities.
Task A focused on interviews with selected PSI personnel and reviews of PSI manuals, procedures, and supporting documents. These interviews and reviews provided the basis for the evaluation and conclusions described herein.
[ An evaluation procedure and checklists were prepared to ensure a thorough and documented review. The checklists were based on requirements
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i in NUREG-0800; 10CFR50, Appendix B; ANSI N45.2; and Chapter 17 of the Pre-liminary Safety Analysis Report (PSAR).
2.2. PSI PERSONNEL AND DOCUMENTS Fifty-nine current and former PSI employees were contacted for this review. Table 2-1 lists 24 of the persons interviewed. The remaining per-sonnel participated in a survey questionnaire, answering prepared ques-tions, providing information, and identifying relevant manuals, procedures, and supporting documents used in the Task A evaluation. Table 2-2 lists the major manuals, procedures, and supporting documents from approximately 120 documents reviewed.
2.3. ORGANIZATIONAL LEVEL AND STATUS OF THE PS1 QA DEFARTMENT The PSI QA organizational level, status, and staffing were evaluated to determine if they are consistent with that required for an effective QA program. The following topics were addressed during the evaluation:
e QA Department staffing.
- QA Department budget.
e Reporting level of QA Manager /QA Cepartment.
- Title of QA Manager compared with peers.
e QA personnel perceptions regarding the stature of the QA Department.
- QA salary levels versus those of counterparts, o QA salary levels versus those of outside organizations.
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2-2
L l
e Organizational independence of QA verifiers from the organizations performing the work.
{
e Training / qualifications of QA verifiers.
- Freedom of the QA Manager and his staff from cost and schedule pressures.
- Freedom of the QA Manager and his staff from non-QA and non-Marble Hill responsibilities.
~
Although no formai staffing criteria exist, the staffing level in the QA Department is based on management experience, familiarity with specific job requirements, and the ability to monitor status of the work. The QA staffing level appears adequate to do an effective job. Based on the evi-dence examined, each QA Department request for staffing increases appears to have been properly considered.
The QA staffing and budget undergo extensive budget analysis, fore-casting, and tracking. The QA Department apparently has adequate indepen-
{
dence and sufficient budgetary control to assure that the necessary man-power and supporting services are planned, obtained, and controlled.
The Executive Director of the QA Department reports to the Senior Vice President of the Nuclear Division and is on a comparable organiza-tional level with the headr, of counterpart organizations.
Staff members in the PSI QA department had varied perceptions of how non-QA personnel felt about them. This is not unusual in the industry.
Overall, the QA Department and personnel appear to have adequate stature for an effective QA program.
L Executive Director is the specific title in the PSI organization for the QA Manager of the QA Department.
[ 2-3
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The salary levels of PSI QA personnel were comparable to counterpart positions in other PSI departments. The PSI QA salary levels also appeared
}
to be competitive with other companies.
QA Department personnel at PSI were found to be organizationally independent from the work being controlled and verified.
Lead auditors were qualified according to ANSI N45.2.23, and inspection / test personnel were qualified according to ANSI N45.2.6.
QA Department personnel were found to be substantially free from
~
excessive cost and schedule pressures. A review of the minutes of 20 Exec-utive Review meetings showed involvement and interest at the highest cor-porate level in assuring a proper balance between quality, cost, and schedule.
No evidence was found of QA personnel involved in non-QA or non-Marble Hill responsibilities.
Shortly before this review began, the PSI QA Officer (the top QA
]
official at PSI) resigned. TPT investiga*.ed the circumstances surrounding his resignation because of possible negative implications to the QA program at MHNGS. TPT determined that the resignation was voluntary and was due to philosophical and personality differences, leading to polarization between the QA Officer and another senior manager. Both individuals reported to the Senior Vice President, Nuclear Division. No specific quality problem or detrimental impact on the QA program was associated with the incident.
No other QA employee resigned or otherwise left PSI at this time. The position of QA Officer was filled by one of the PSI QA managers. The posi-tion title was changed to Executive Director of Nuclear QA, with no change in authority, responsibility, or reporting level.
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The organizational level and status of the PSI QA Department were found to be satisfactory and consistent with that required for an effective
{ QA program. The QA Department is satisfactorily independent and has suffi-cient authority and stature to do an effective job.
No PFR was written against this portion of Task A.
2.4. AUTHORITY AND RESPONSIBILITY OF QA PERSONNEL Job descriptions and procedures relevant to defining the authority and responsiblity of key PSI QA personnel were reviewed. These documents were evaluated to determine if the identified responsibilities and authorities were commensurate with those required for an effective QA program. The following topics were addressed during this review:
e QA Manager position description and qualifications.
- Key QA personnel position descriptions and qualifications.
e Authority delegated by upper management to QA Manager.
e Other vehicles / procedures which define QA authority.
- Freedom in QA Department's use of stop-work authority.
{
A review of job / position descriptions of four selected managers in the QA Department and five other key QA employees indicated that the duties and responsibilities of each position examined were not adequately defined at
[ the time of the review. Hotiever, new job descriptions, currently in the review cycle at PSI, are adequate.
A review of several manuals, procedures, and statements of authority indicated that the authority and responsibility of the QA organization are well defined and appropriate. The QA Officer has been assigned responsi-bility for and authority to implement the QA program.
QA Officer is the former title for Executive Director of the QA Department.
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l The QA Department has free use of stop-work authority without hindrance from others (e.g., all personnel performing QA functions can limit or control further processing or installation of an item). The Senior Vice President, Nuclear Division, has a responsibility to assure that the work is stopped immediately and is not allowed to modify the Stop Work Order. The signatures of both the Senior Vice President and the QA Officer are required to lift the Stop Work Order.
No valid PFR resulted from this portion of Task A.
2.5. QA DEPARTMENT ACCESS TO UPPER MANAGEMENT The PSI QA Department's access to upper management was evaluated to determine if the QA program status and problems can be, and have been, brought to the attention of upper management and acted upon, as appropri-ate, in a timely manner. The following topics were addressed during the evaluation:
e Methods for the QA Manager to report QA program adequacy and effectiveness to management.
e Other vehicles for informing management of quality status / problems / trends:
Corrective Action Requests.
Nonconformance Reports (NCRs).
Progress reports.
}
Audit reports. '
Memos.
Trend reports.
Other reports.
e Distribution of contractor audit reports, trend reports, etc., to PSI management.
s 2-6
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( e Contractor reporting of problems to PSI management.
e Management response to reports and QA program status / problem /
information received.
e QA personnel perceptions regarding an "open door" policy for reporting program status and problems to management.
e QA Manager's attendance at upper management meetings.
J e Extent, if any, of censoring of information on problems
. as reports proceed up the management chain.
The QA Department was determined to have adequate access to upper management, and QA program status and problems were found to be regularly brought to the attegtion of upper management.
The QA Executive Director attends the monthly Executive Review meet-7 ings, which are also attended by all division upper management through the Senior Vice President, Nuclear Division, and the PSI President and Chief L Executive Officer. As evidenced by the meeting minutes, these meetings
, have provided a formal mechanism for QA matters to be brought to the atten-tion of top management at PSI. In addition, the QA Executive Director is in close daily contact with the Senior Vice President, Nuclear Division.
( The QA Executiva Director attends other project meetings, including bimonthly staff meetings held by the Senior Vice President. The President of PSI also generally attends the Senior Vice President's staff meetings.
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Thus, excellent access and adequate communication is assured on QA matters.
Quarterly Quality Trend Reports provide another mechanism for inform-ing management of quality trends and status. These reports are distributed to management levels up to the Senior Vice President, Nuclear Division.
2-7 f
A majority of PSI QA and Nuclear Services staff members felt that PSI did, indeed, have an "open door" policy for reporting problems to management and that PSI management has, in fact, been responsive in acting on problems in a timely manner.
PSI upper management has been found to be aware of the implementation of the QA Program. _
It is unlikely that they have been or will be insulated from the actual project status and/or problems.
No valid PFR was initiated on this portion of Task A.
2.6. QA DEPARTMENT INVOLVEMENT IN PROJECT ACTIVITIES An evaluation was performed to determine if the PSI QA Department was sufficiently involved in project activities to help assure adequate control 4 and cognizance with respect to project quality. The following topics were addressed during the evaluation:
o PSI QA Department reviews and audits of contractors for adequacy, implementation, and effectiveness of their QA programs.
s e Timeliness and extent of PSI QA Department surveillance of construction.
PSI QA Department review of contractor inspection planning.
PSI QA Department identifict '-f >c and use of witness and hold points.
PSI review of contra g > O ...entation.
I e PSI QA Department interfaces with contractors and the NRC on matters of urgency.
I e PSI QA Department involvement in the responses to the NRC.
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e PSI QA Department inspection on equipment turnover.
e PSI QA Department concurrence on quality-related procedures, including engineering and construction procedures.
An examination of the site and supplier annual audit schedule verified that the site contractors and suppliers were audited on a regular basis to appropriate criteria. Selected audit reports of site contractors that were reviewed showed that their QA programs had been evaluated by PSI QA for adequacy and effectiveness and that, as necessary, corrective action had been taken on shortcomings reported during the audit.
A review of the PSI QA Department mandatory hold point logs and con-struction' surveillance reports with related documentation indicated that the PSI QA Department reviews contractor inspection planning and procedures to establish witness and hold points. The necessary controls are and have
- been in place to assure adherence to the hold points. A review of the Corrective Action Requests, PSI / contractor / supplier NCRs, requests for evaluation submitted to the Safety Committee, and other problem-reporting documentation on matters of urgency are adequate. Furthermore, the QA Department appears to be appropriately involved in the preparation and/or
(
submission of the responses to NRC inspection reports.
A review of the minutes and attendance of the Project and Construction Department status and problem-solving meetings revealed that the QA Depart-ment participates and plays an active role in these meetings.
An examination of a number of PSI project and engineering procedures, site contractor construction / installation and test procedures, and PSI QA 2-9
1 I
procedures indicated that the appropriate QA disciplines reviewed and approved the procedures.
No valid PFR was initiated on this portion of Task A.
2.7. MANAGEMENT INVOLVEMENT IN QA ACTIVITIES PSI management involvement in QA activities was evaluated to determine if management was sufficiently involved to provide an appropriate level of support and status to the QA program. The following topics were addressed curing d:e evaluation:
e Management's awareness of problems.
s Management's response to quality-related requests for support.
e Frequency of management's contact with QA program status.
e Management's stated policies regarding quality /QA.
e Management's statements expressing an opinion on QA.
e Management emphasis on " Quality is Everybody's Business" and "Do g It Right the First Time" themes. I e QA audits and evaluations requested by management.
e Employees' perception of management's policy on quality /QA.
. E.p1o,..s. p.rc.ption o, .anag. ment's attitude toward ana commitment to QA.
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h The review showed that PSI's management is, in fact, closely involved J ~ with QA activities and does provide an appropriate level of support and L status to the QA program. The Executive Review meetings involve PSI top management (Chairman and President) on a monthly basis, and the Senior Vice President, Nuclear Division, is involved daily through his contacts with the QA Executive Director. Interviews with QA personnel from line posi-a tions through middle and upper management all indicated strong support of
[
the QA program by PSI mansgement.
I A PSI policy statement in support of the QA program appears in the QA manual, signed by D.V. Menscer, President.
y PSI training and indoctrination programs for new employees stress that PSI management supports the QA program, and also stress the " Reporting Without Discrimination" policy.
Management audits are called for in the Project QA Manual (PQAM) and the American Society of Mechanical Engineers (ASME) QA Manual (AQAM). A review of management audit files for the last three years confirmed that two audits were performed each year (one for the PQAM and one for the AQAM). However, a review of the contents of these reports indicated that they were limited, in TFT's opinion, to implementation audits and did not adequately evaluate the effectiveness of the program, and one audit did not assess compliance with codes, standards, and Regulatory Guides. PFR 2485:011 was written on this issue and classified as a Finding.
PSI is in the process of issuing written instructions for performing management audits. Prior to this, an uncontrolled document signed by S.W.
Shields, " Instructions for Performance of Management Audits at Marble Hill Nuclear Generating Station," November 5,1981, apparently served as the governing procedure. Although the management audit files did not always contain all the backup records required, the basic reports and essential supporting records were available.
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Interviews with D. V. Menscer (President and Chief Operating Officer of PSI) and H. A. Barker (Chairman and Chief Executive Officer of PSI) indicated a strong and sincere commitment to maintaining an effective QA program throughout the project. Both Mssrs. Menscer and Barker believe that a strong QA program is essential for the successful completion of the project and for the safe and reliable operation of Marble Hill. They encourage reporting of any problems related to QA and the prompt investi-gation and assessment of any allegations reported by site personnel. In addition, they encourage prompt, frank, and open communication with the NRC inspector.
The PSI QA and Nuclear Services staff felt that PSI management had a positive attitude toward and strong commitment to QA.
One valid PFR (2485:011), classified as a Finding, resulted from this task. The Finding addressed management audits of the QA program. The I reviewed audits, in TPT's opinion, did not adequately assess the effective-ness of the QA program, and one audit did not assess compliance with codes, standards, and Regulatory Guides.
A CAP was formulated by PSI to address the concerns on the Finding.
The CAP requires a change (which has been effected) in project management a procedures to ensure that a more substantive statement is made about QA program effectiveness and that compliance to codes, standards, and Regula-tory Guides is checked in the future. A check for similar problems was not I necessary, since all management audits of QA made over the past three years were reviewed by TPT.
2.8. CONCLUSIONS, TASK A Based on the results of this evaluation and the correction of the one isolated deficiency, PSI management is concluded to have done a good job of developing and implementing policies which are supportive of the Marble Hill QA program. Overall, the PSI QA Department appears to receive 2-12
L excellent support from management. The QA Department is actively involved in the surveillance of the Marble Hill QA activities and appears to have sufficient authority to carry out its responsibilities. A key element in the success of the Marble Hill project since 1980 is the extensive involve-ment and sincere interest in quality and the QA program on the part of upper and top management.
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- TABLE 2-1 PERSONNEL INTERVIEWED IN TASK A(,)
E. Aimone Division Personnel Manager D. Albright Senior Quality Engineer, Electrical H. Barker Chairman and Chief Executive Officer C. Beckham Executive Director, Nuclear Quality Assurance J. Bott Manager, Nuclear Regulation and Affairs F. Carchedi Mechanical Quality Engineering Manager E. Carmichael Senior Quality Engineering, Mechanical H. Curry Electrical Quality Engineering Manager D. Dedrick Audits Manager C. Driskell Senior Quality Assurance Engineer l T. Geyman Civil Test Laboratory Supervisor l S. Gordy Qualifications and Training Administrator A. Heestand Training Department g J. Huffman T. Kunze Senior Quality Engineering, Electrical Personnel Manager l
J. Lefman (Acting) Quality Administration Supervisor M. Macy NCR Coordinator D. Menscer President and Chief Operating Officer R. Minnich Senior Quality Engineer, Mechanical B. Morrison Quality Engineering Manager g W. Petr Vice President Nuclear Services ]
L. Ramsett(b)
S. W. Shields Senior Vice President, Nuclear Division J. Stivers Project Review Board Chairman
- In addition, the site NRC representative, J. Schapker, was also contacted.
(b)Former PSI employee, QA Officer.
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TABLE 2-2 DOCUMENTS REVIEWED IN TASK A Manuals and Procedures Project Man'agement Procedures (PMP)
Section 10 2-3, Rev. 3 2-4, Rev. 0 5-3, Rev. 3 5-4, Rev. 1 7-7, Rev. 1 10-2, Rev. 2 16-3, Rev. 2 Project QA Manual (PQAM)
Section 1, Rev. 9 Section 2, Rev. 10 Section 15, Rev. 9 Section 16, Rev. 8 PAQM Policy Statement 7/16/82, 7/27/81, 6/4/80, 1/22/80 Project Management Manual, v. 1 and 2 Section 16.2 PPM Policy Statement J ASME QA Manual (AQAM) 3 Statement of Policy and Authority, 7/16/82 Statement of Authority, 10/22/82 Section 2, Rev.11 1 Test Methods Procedures Manual Calibration Procedures Manual Civil Inspection and Testing Service Manual 2-15
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TABLE 2-2 (ccatinued)
+
QCP-5 Rev. 12 TMP-15 Rev. 5 CAL-12 Contractor Procedures Cherne 15.4, Rev. 12. " Installation of Mechanical Equipment" Cherne 11.1 !
CONAM 59-MT-023 Newberg QAP 9.01 l
Pullman PWP 2.2 Commonwealth-Lord QWP-C11 Sargent & Lundy Engineers PI-MH-004, Rev. K Reports / Meetings Executive Review Meeting Minutes, January 1981 to June 1983 Quality Trend Reports 1982 Quarterly Reports 5/16/82, 9/1/82, 11/12/83, 3/18/83, 5/11/83 Contractor Trend Reports Quality Engineering Monthly Report 6/30/83 and Weekly Reports prior to that date Construction Surveillance Reports No. Cherne CC-083-0339 and Cherne CC-083-093 completed 7/15/83 Safety Review Committee meeting minutes No. 409, 2/26/82 Suspected Trend Investigation Report (STIR), 2/8/83 Project Review Meeting notification memo and agenda, 5/16/83 and 5/25/83 Salary and Budgets Budgets for 1981, 1982, and 1983 8 1983 QA Department budget preparation sheets Budget estimate for QA Department, 8/81 and 3/82 Budget Variance Reports 1981, 1982, and 1983 2-16
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TABLE 2-2 (continued) 1983 Compensation Administration Program 7/19/83
. Salary Information from E. Almone Training Qualification and Training Folders Level A Orientation
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Lesson Plan, Rev. 3 Shields, " Introduction for Performance of Management Audits," 11/8/81 Staffing and Job Descriptions Actual staffing levels from November 1981 Manloading Charts for QA Department from January 1981 to present Job Descriptions Operations Quality Assurance Supervisor Senior Operations QA Engineer Operations QA Engineer Senior QA Engineer Quality Administration Supervisor Quality Engineering Manager Senior Document Inspector Level III Senior Quality Engineer Quality Engineering Surveillance (Lead)
Quality Engineer Quality Assurance Receipt Inspector Supervisor Quality Records Verification Manager I Discipline Quality Engineering Manager Supplier Audit Supervisor Senior Supplier Quality Engineering Supplier Quality Engineering Representative Site Audit Supervisor 2-17 f
1 TABLE 2-2 (continued)
Audit Manager Senior' Quality Assurance Engineer (Lead Auditor)
Startup Quality Assurance Engineer Startup Quality Assurance Manager Senior Startup QA Engineer Logs Stop Work Order Status Log Quality Engineering Mechanical Mendatory Hold Point Status Log for j Cherne s Quality Engineering Electrical Mandatory Hold Point Log for Commonwealth-Lord l k
Audits Site Annual Audit Schedule, first quarter update, 5/12/83 Supplier Annual Audit Schedule, first quarter update, 5/10/83 Cherne (Site) Audit 83-Cherne-01, completed February 1983 Nuclear Regulatory Commission Audit 81-22, 4/13/82 (Docket No. 50-546, 50-547) 83-MGMTAUD-01 (ASIIE), 3/83 82-MGMTAUD-02, 12/82 82-MGMTAUD-01 (ASME), 5/82 NUS-3950, 12/81 81-MGMTAUD-01 (ASME), 5/81 16-MGMTAUD-01 (ASME) 16-MGMTAUD-02 (NUS), 3/81, Rev. 1/81 Stop Work Orders, Corrective Action Request (CARS), and Nonconformance Reports (NCRs)
All Stop Work Orders listed in Stop Work Order Status Log Managment CARS I-83-01, I-83-02, I-83-03, A-83-01, A-83-02 CAR 282 PSI 0047, 2/25/82 CAR 283 PSI 0040, 3/7/83 Commonwealth-Lord NCR 54L8 CAR 282 PSI 220 2-18
TABLE 2-2 (continued)
Miscellaneous PSI Organizttien Charts Quality Engineering letters to Conunonwealth-Lord establishing Marble i Hill Procedures Request for Evaluation (RFE) S-82-04, 2/25/82 and RFE S-83-09, 3/9/83 Critical Items List meeting notification memo and action item sheets for 7/8/83 2-19
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' 3. CONSTRUCTION DESIGN CONTROL, TASK B 3.1. TASK OBJECTIVE AND SCOPE l
This task reviewed and evaluated the PSI Marble Hill system for control of approved design documents, SSL-initiated design changes, and field-initiated design changes, and evaluated implementation of the system and its effectiveness in assuring that safety-related components are constructed in f accordance with the approved design.
The scope of this task included the following:
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e Review and evaluation of the PSI Marble Hill system and procedures for control and issuance of approved design documents, design changes, and field changes against the requirements of 10CFR50, Appendix B, and ANSI N45.2.11, 1974.
L e Review and evaluation of a selection of design documents, change noticea, and field change documents for compliance with applicable
[ procedures.
I Working procedures and checklists were prepared for this task to ensure a thorough and uniform method of performing the evaluation by TPT personnel.
~ The checklists also provide a detailed record of the work.
L g 3.2. . DESIGN CHANGE CONTROL PROCEDURE REVIEW L
The relevant construction design control manuals and procedures were I
obtained from PSI. Table 3-1 lists the manuals and procedures obtained.
These procedures were reviewed to determine whether the design control l
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requirements listed in 10CFR50, Appendix B, and ANSI N45.2.11-1974 were adequately addressed in working-level procedures and manuals.
The procedure review consisted of the f ollowing:
- 1. Identifying and extracting the design control requirements in 10CFR50, Appendix B, and ANSI N45.2.11-1974.
- 2. Entering the requirements as checklist items on a specially prepared checklist form.
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- 3. Examining the manuals and procedures in detail and recording on the checklist the specific section(s) in which the requirement was addressed.
- 4. Indicating on the checklist whether or not the requirement was adequately addressed. When judgment had to be exercised as to adequacy, comments were added to justify the reviewer's decision as to adequacy.
Specific details for meeting QA requirements are described in the PSI Project Management Manual as Project Management Procedures (PMPs). For construction design changes, the most important are as follows:
PMP 3-1: Design Document Review.
PMP 3-2: Field Change Requests.
PMP 6-1: Document Control.
In addition, the following S&L Project Instructions (PIs) are important in construction design change control:
I PI-MH-005: Preparation of Drawings.
PI-MH-009: Preparation of Specifications.
PI-MH-017: Site Design Control.
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PSI is responsible for surveillance of contractors / subcontractors to ensure that QA objectives are met. Westinghouse, the NSSS designer, and S&L, the balance of plant system (BOP) designer, are responsible for surveillance of their own specific subcontractors. The design of MHNGS Units 1 and 2 is essentially a duplication of the Byron Units 1 and 2. Many of the engineering documents are copies of Byron documents. S&L is the archit=ct/ engineer for both MHNGS and the Byron units and is responsible for verifying replicate and nonreplicate features for Units 1 and 2 of S&L- and supplier prepared designs. S&L is also responsible for identifying the k external design interfaces on the Marble Hill project.
The principal vehicles for construction design cha:.ges are Field Change
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Requests (FCRs), which originate at the MHNGS site, and Engineering Change Notices (ECNs), initiat.ed by S&L, Chicago. FCRs and ECNs are used to revise drawings. However, only ECNs are used to revise specifications, incorporate
- design changes, and make improvements over the base plant design. FCRs are initiated by PSI, often at the request of a construction contractor (e.g.,
Cherne), usually because of minor interference problems encountered during construction. The PSI Resident Engineer requests S&L Site Engineering to revi*e the request for acceptability. If the proposed change impacts the base design significantly, the FCR is sent to S&L, Chicago, for more extensive review concerning the impact of the change on other systems and structures before the proposed change is implemented. Because interference problems requiring changes can delay construction work, FCRs sent to Chicago are rapidly expedited for quick approval and returned to the Marble Hill site. Review of the processing of ECNs and FCRs by S&L was not part of the scope of this review. However, the processing of an ECN or FCR af ter release by S&L to the site was reviewed.
Multiple copies of revisions of documents by S&L, Chicago, are sent to the PSI Site Document Control Center for distribution to the PSI internal
. organizations and external PSI contractors / subcontractors.
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l Two types of review of design documents and changes are performed by PSI Project Engineering: the Acceptance Review and the more extensive i Engineering Release Authorization (ERA) review. A Document Review Notice (DRN) accompanies the ERA and is sent to the appropriate reviewers as required by PHP 3-1. These design documents are not distributed until the review has been completed.
Control logs are maintained by the Site Document Control Center and by Project Engineering to track design documents and to prevent release of unapproved design documents. In addition, PSI has a computerized Design Document Information System (DDIS) with interactive computer terminals in many locations at the Marble Hill site and at external contractor office locations. Design document data are entered into the computer files by authorized, trained personnel and can be accessed by anyone at a terminal.
Only current outstanding changes against a design document are displayed.
Changes already incorporated in design documents were deleted from the active files.
In general, the PSI system for controlling construction design changes to safety-related components meets the requirements specified in 10CFR50, I Appendix B, and ANSI N45.2.11. One exception to compliance with QA require-ments was documented in PFR 2485:005 and pertained to design review respon-sibilities. After a review of PSI design review procedures, it was not clear as to what role PSI played in the design review process. A discussion with PSI indicated that the present revision of the Design Control Procedure included inappropriate references to " design reviews," that these references had been identified, and that a revision to the procedure to eliminate these references would be issued shortly. This PFR was classified as an Observation since PSI had delegated design review to S&L and its procedures I were judged to be adequate in this area.
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3.3. PROCEDURE IMPLEMENTATION REVIEW The PMPs were the principal sources for composing checklists to review
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the construction, design documents for implementation of design change con-trol procedures. Some S&L PIs were used as necessary supplements to ensure that construction design documents met QA requirements. Checklists were developed on the following:
- 1. Drawings: content, review / approval, field-initiated changes,
[ tracking.
- 2. FCRs: content, review / approval, revisions.
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- 3. Specifications: content, review / approval, revisions.
< 4. ERAS: identification of reviewers required, type of document, tracking.
Documents selected from those features covered in the walkdown in Task C, consisting of 97 drawings, 59 FCRs, 8 Specifications, and 18 ERAS, were reviewed during the implementation review. Table 3-2 lists the design documents reviewed.
No valid PFR was initiated on this portion of Task B.
3.4. CONCLUSIONS, TASK B Based on the review performed in Task B, the construction design control system and procedures are concluded to be adequate and properly implemented. ,
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- TABLE 3-1 PROCEDURAL DOCUMENTS REVIEWED IN TASK B Designator Title 1
PQAM Project Quality Assurance Manual AQAM ASME Quality Assurance Manual PMP 3-1 Design Document Review PMP 3-2 Field Change Requests PMP 6-1 Document Control PI-MH-005 Preparation of Drawings PI-MH-009 Preparation of Specifications PI-MH-017 Site Design Control
- PMPs denote Project Management Procedures and are PSI documents. PI-MHs denote Project Instructions for Marble Hill and are S&L documents.
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3-6 -
TABLE 3-2 DESIGN DOCUMENTS REVIEWED IN TASK B System /
Structure -
Drawing No. ERA No. FCR No.
Containment S-919/AW 4790 C-8298 Building -
A-6967 -
L S-953/K -
C-8298 21T 938-938-26/2 - -
21T 938-938-28/2 - -
21T 938-938-31/3 _
1098-E919AB/E - --
10981-1022/- - --
[ 10981-1098/B 10981-1176/D 10981-1194/C - --
h S-885/P 4357 C-7966 EW-S-321 -
S-876/AK -
C-8298 A"n 11ary S-1332/N 4806 -
Building 10982-E-1332/E - -
10982-E-4083/B - --
10982-E-4178/B - --
c 10982-E-4180/C - --
' S-1286/W -- --
r CCWS M-66-IS 3764 M-5790 M-6475 M-66-2J -
M-6532
' M-66-3M 3542 --
[~ M-537-5L --
C-8449
' M-7153 M-547-8J -
M-537-9P -- --
r' M-537-15K --
M-6023 L M-537-18E - --
M-537-20L --
M-6565 r M-537-22H - --
M-537-24H - --
M-537-25F -
M-6912 M-6357 Piping Line List, -
M-7135 Rev. 31 1CC04012R, Rev. B -
M-6711 1CC04014X, Rev. C -
M-4823 1CC04015X, Rev. B - --
e 3-7 L
E
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l TABLE 3-2 (Continued)
Structure Drawing No. ERA No. FCR No.
CCWS 1CC04016R, Rev. C - -- 5 (Continued) 1CC04034G, Rev. B - --
W-1CC40014X, Rev. D -
M-8014 l W-1CC40015X, Rev. D -
M-8007 l 3-1CC40021X,Rev.B --
M-9246 W-1CC40026X, Rev. C - --
W-1CC40017X, Rev. D M-9344 Fuel Handling S-812/Z 4193 -
Building S-817/P - -
S-818/E -- --
S-819/F -
C-7887 S-822/K --
C-7103 xx l S-1144/g -- --
l S-1416/g - --
S-1418/M --
C-7887 S-476/V -
C-7453 I Ultimate Heat Sink S-373/H --
C-7390 S-374/E -- --
S-381/R --
C-5188 C-7035 S-382/1 -
C-5188 S-383/T 5088 C-7824 S-389/M --
C-7964 Miscellaneous S-379/G -
C-6391 drawings used in S-384/G -
C-6650 Subtask D2 S-387/AE --
C-6391 C-7227 S-497/AM 4520 -
S-1107/T -
C-7366 S-1606/N - -
S-1610/Y 4528 C-8384 C-8385 S-1617/V -
C-7533 C-8005 M-60-1K RCS M-60-4K lg M-60-5J -- --
l M-542-1H -
M-7713 M-542-2J -- --
M-542-3E -- --
M-542-4J -
M-6035 M-542-5J -- --
3-8 j
[
s TABLE 3-2 (Continued)
System /
[ Structure Drawing No. ERA No. FCR No.
7 RCS (Continued) M-542-6J - --
[ M-652-7P -
M-6356 M-7886 M-652-8M -
M-6355
[ M-1RC04001V - --
M-1RC04002S - --
M-1RC04003S - -
( M-1RC04004V - --
AFS M-37-1K 3497 --
M-541-1Z -
M-5213 M-5276 M-541-4J - --
[ M-541-8K -
M-5112 M-541-10G -
M-5147 W-9518D23 - --
[ PP-300BN50366, -- -
Rev. 3, 4
( RHR M-61-4M M-52-1R 3878 3757 M-8250 M-538-1K -
M-5829 M-538-2T -
M-8250
[ M-538-3W -
M-6414 M-538-4R -
M-6520 M-538-5M -
M-5829 W-5874, Rev. 3 Mechanical piping Y-2739, Am. 10, 3814 M-4621
[ Am. 8 (ECN-M-562)
Y-2741, Am. 6, 3998 -
Am. 5 Y-2721, Am. 9 4991 --
Y-2722, Am. 20 4082 --
Y-2850, Am. 17 5089 -
{ Y-2944, Am. 17 4081 --
[
[
3-9
[
f - - - - -
[
[
[
. 4. PHYSICAL VERIFICATION-WALKDOWN, TASK C 4.1. TASK OBJECTIVE AND SCOPE This task determined if the physical installation and construction of selected portions of safety-related mechanical systems and structures of
(- MHNGS conform to the requirements of design drawings and specifications.
Due to the construction stage of MHNGS Units 1 and 2 (currently approx-
{
imately 50% and 25% complete, respectively), the scope of this task was limited to verification of safety-related mechanical systems and structures which were furthest along in completion status. Electrical and instrumenta-tion syatens were not reviewed.
The scope of this task included the following:
[
e Prepare the mechanical and structural verification procedures.
- Select the mechanical systems and structures to be physically verified.
e Perform the detailed mechanical and structural verification.
The physical verification was conducted primarily on Unit i features.
- Approximately 25% of similar features were verified on Unit 2.
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4-1 f - - - _ - - - - - -
4.2. FEATURE SELECTION Mechanj: 11 and structural features to be physically verified were selected based on the following criteria:
- 1. Features supplied, installed, or constructed by major contractors (e.g., Westinghouse, Newberg, and Cherne).
- 2. A range of sophistication in construction methods, including in-process construction activities.
- 3. Features to include:
I
- a. At least one safety-related mechanical / structural component.
- b. At least one major NSSS component.
- 4. A wide range of equipment, structural types, and complexity to provide balance in the review.
- 5. Portions of the systems or structures including safety-related components such as piping, pipe supports, an electrical raceway, reinforcing steel, mechanical splices, concrete anchors, and a structural frame members.
l
- 6. Features which are encompassed within one or more identified concerns of the NRC.
I A feature is a system, subsystem, component, or structural type which is a definable whole or part of a whole. Examples of features are the Resi-dual Heat Removal System (RHR), a steam-turbine-driven pump, a building, a portion of a building, pipe supports, reinforcing steel, concrete placement, valves , etc.
4-2 -
( Four safety-related systems were chosen from which mechanical features were selected for verification: the Reactor Coolant System (RCS), the Auxiliary Feedwater System (AFS), the Component Cooling Water System
(
(CCWS), and the Residual Heat Removal System (RHR). These systems were the most advanced in completion status at the time of the review. The systems also contained a representative cross section of mechanical features meet-ing the selection criteria. Figures 4-1 through 4-5 show the portions of the systems in Units 1 and 2 selected for verification. Tables 4-1 through 4-7 give detailed lists of the selected components. Section 4.4.1 gives
[ additional descriptions.
Five safety-related structures were selected for verification: the
(
Reactor Containment Buildings (Units 1 and 2), the Auxiliary Building, the Fuel Handling Building, and the Ultimate Heat Sink structure. These struc-tures were considered representative of the different types of structures at MHNGS. The Reactor Containment Building is a post-tensioned reinforced concrete cylindrical structure; the Auxiliary and Fuel Handling Buildings are of the structural steel frame and shear-wall type of construction, and the Ultimate Heat Sink structure is a reinforced concrete shear-wall type of building. The status of construction of the various buildings at the
( time of review allowed both completed structures and in process construc-tion activities to be verified. Figures 4-6 through 4-10 show the portions of the structures in Units 1 and 2 selected for verification. Tables 4-8 through 4-12 give detailed lists of selected structural features. Section 4.4.2 gives additional descriptions.
[
4.3. WALKDOWN PROCEDURES
[
Separate procedures were prepared for use in verifying mechanical
( systems and structural features. The procedures established a uniform and comprehensive method for performing the walkdown task.
[
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4-3
{
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l The procedures specified three levels of verification as follows:
1 Level 1: Verify overall system installation as defined by the systems i piping and instrumentation diagrams and major safety structures as depicted in the structural layout drawings and specifications.
Level 2: Verify mechanical and structural component installation in terms of identification, location, orientation, and configuration, as depicted by the design drawings, specifications, and related equipment documents.
l l Level 3: Verify installed details of the mechanical systems and structures, such ac detailed dimensional checks of piping and pipa supports, details of structural members, welds and bolt torques at connections, details of reinforcing tendon anchors, reinforcing steel details, and concrete placement.
The procedures provided detailed instructions for preparing walkdown packages which include applicable design documents for verifying the fea-ture. A supplemental set of forms was used as checksheets listing specific I items to be checked and to record data collected during the walkdown.
l The procedures also provided criteria for determining whether a noted I deviation was a Potential Finding.
4.4. PHYSICAL VERIFICATION PERFORMANCE The physical verification was performed separately for (1) mechanical systems and (2) civil structures. E I
4-4
L I l r
4.4.1. Verification of Mechanical Systems The physical verification of mechanical systems was organized into 52 j walkdown packages (44 for Unit I and 8 for Unit 2). These packages con-sisted of 87 levels of verification (i.e., 5 level 1, 61 level 2, and 21 level 3). The verification at levels 2 and 3 covered 251 items of piping, El pipe supports, and mechanical equipment. A complete set of documents was l
s accumulated for each package upon which to base the details of the physical verification. The documents included P& ids, single system drawings (iso-metric drawings), FCRs, ECNs, valve lists, equipment lists, pipe support and piping lists, plant and engineering installation drawings, area draw-ings, and equipment drawings. Where applicable, ASME Code data sheets and manufacturer's drawings and reports were also included. The PSI computer-based Design Document Inforration System (DDIS) and Materials Management Information System (MMIS) were frequently used to obtain design change history. More than 630 individual documents were consulted or otherwise referenced during the preparation of the various walkdown packages.
A set of forms was prepared for each mechanical walkdown package to list each of the piping, pipe support and/or equipment items. These forms then became checksheets, and any notes pertinent to the items being checked could be included therson. These forms were extremely useful as detailed listings of precisely what was examined during the walkdown.
The features selected for physical verification were representative of a cross section of mechanical components and equipment. Approximately 1800 linear feet of installed piping (including 420 linear feet of Unit 2 pip-
{
ing) with diameters ranging from 3 to 31 in. and 67 installed pipe supports of various types were included in the walkdown. In addition, 50 pipe sup-ports were determined to be partially installed, 89 not yet installed, and two inaccessible for verification. Table 4-1 summarizes the installed pipes and Table 4-2 summarizes the pipe supports verified in the mechanical walkdown. The physical verification of major equipment included one steam 4-5
- _ _ _ _ _ _ _ _ _ _ . i
i generator, one pressurizer, five pumps, seven heat exchangers,16 motor-operated or air-operated valves, 54 other types of valves, two diesel engines, and one electric motor. Table 4-3 summarizes the mechanical features verified for Units 1 and 2. The following paragraphs briefly describe portions of the four safety-related systems subjected to verification.
4.4.1.1. Reactor Coolant System. Figure 4-1 shows the portion of the RCS selected for physical verification for Unit 1. One of the four reactor coolant loops was selected for system verification at level 1. Tables 4-1 and 4-2 list the RCS piping (stainless steel) and pipe supports verified at levels 2 and 3.
I Table 4-4 summarizes the RCS equipment verified at level 2. These components and equipment included one reactor coolant pump; a Westinghouse vertical, single-stage, shaf t-seal, centrifugal pump; one Westinghouse ver-tical U-tube design steam generator; an electrically heated Westinghouse-supplied pressurizer; three motor-operated gate valves; one relief valve and two rechanical gate valves.
4.4.1.2. Auxiliary Feedwater System. The AFS consists of two independent and redundant, 100% capacity trains. One train uses an electric motor- a driven pump and the other uses a diesel-engine-driven pump with a speed increaser. The major AFS components include a 10-stage horizontally mounted centrifugal motor-driven pump, a diesel-engine-driven pump, essen-tial service water supply valves, discharge test valves, recirculation iso-lation valves, flow regulation valves, and isolation valves.
Figure 4-2 shows the portion of the Unit 1 AFS selected for physical verification at level 1. Eleven major AFS components were verified in Unit 1; one piece of AFS equipment in Unit 2 was included. Some AFS components were verified in storage in the warehouse at the site. Table 4-5 lists AFS 4-6
L equipment and components that were verified at level 2. Tables 4-1 and 4-2 list AFS piping and pipe supports that were verified.
j 4.4.1.3. Component Cooling Water System. Units 1 and 2 share the CCWS.
Each unit has three major paths for the component cooling water from the l heat exchanger outlet. Two paths lead to separate RHR cooling trains. One path was selected for the physical verification of piping, components, and pipe supports for Units 1 and 2. The third path, (through MOV9415) leads to the normal CCWS heat loads that are normally cooled by the CCWS through (I branch headers that are at the top of the main header and downstream of valve MOV9415. These include the RCS supply and return header, excess let-down heat-exchanger-loops, the pressurizer sample cooler loop, letdown heat exchanger loops, and the spent fuel pit heat exchanger loop. A representa-tive portion of each subbranch loop mentioned above was included in the physical verification for Unit 1.
Figures 4-3 and 4-4 show the portions of the CCWS for Units 1 and 2, respectively, that were physically verified. Fif ty-three major components were selected, including one pump, four heat exchangers, seven motor-operated valves, five check valves, and a variety of other valve types.
Twenty-one of the fifty-three components were not yet installed. Table 4-6 summarizes equipment and components included in the physical verification for both units.
Table 4-1 summarizes the CCWS piping included in the physical verifi-cation. Piping ranged in diameter from 3 to 18 in. Most of the pipe lengths verified were in the CCWS as were a large proportion of pipe sup-ports verified. Table 4-2 summarizes the installed pipe supports that were physically verified.
4.4.1.4. Residual Heat Removal System. Each unit has two redundant RHR flow trains, and each train consists of a pump, a heat exchanger, flow con-trol and motor-operated valves, and the associated piping. Figure 4-5 shows the portion of the RHR in Unit 1 included in the walkdown. Thirteen 4-7
major RHR componencs were selected from Units 1 and 2 for verification, including one centrifugal pump (590 psig, 500 gal / min), two heat exchangers (tube and shell type), four air-operated valves, three motor-operated valves, and three other types of valves. Table 4-7 lists the RHR equipment :
and components included in level 2 verification. Tables 4-1 and 4-2 sum-marize the RHR installed piping and pipe supports that were verified.
Walkdown of the selected portions of the mechanical systems at level i verified if the mechanical components and piping are installed in their proper sequeace, as shown in the piping and instrumentation diagram. It also checked that all system components appear on the design diagrams.
Level 2 and 3 verification included the following checks:
o Pipe diameters conform to design requirements.
e Pipe routing of large bore pipes conform to area drawings and oculine system drawings (piping isometrics).
o Pipe and equipment support location, orientation, and configura-tion conform ta design documents such as piping isometrics and pipe. support drawings.
e Equipment nameplate data correspond to equipment specification u
requirements.
Equipment tag numbers match equipment list designations, e Detailed dimensional tolerances of components and component sup-ports, including material welds and connections to supporting structures meet the design requirements.
4-8
k
{ As a result of the physical verification of selected Unit 1 and 2 mechanical systems, involving more than 4500 individual checks, two valid PFRs (classified as Observations) were issued as follows (see Table 7-1).
PFR 2485:007, classified as an Observation, recorded a deviation found in the location of a rigid pipe restraint, tag No. 1CC03052R. The pipe sup-port is used for supporting a 16-in.-diameter pipe, tag No. ICC05G16, of the CCWS located in the Auxiliary Building. The as-built pipe support is located 12 in. from the specified design location, thus deviating from the
- 6-in. installation tolerance. Investigation by TPT, including an indepen-dent stress analysis of the pipe and pipe support in their as-built loca-tion, determined that the as-built location does not cause pipe or support stresses to exceed ASME allowable stresses. Pipe support 1CC03052R and the other supports installed prior to June 1982 will be reverified by Cherne; an NCR at that time would be subject to engineering disposition.
PFR 2485:016, classified as an Observation, pointed out an inconsis-tency between the design specification and the installed configuration of several (12 identified) safety-related valves of the Units 1 and 2 CCWS.
Further investigation determined that the required changes involving addi-tion of valve position-indication switches were incorporated by Westinghouse under PSI design criteria DC-RN-01-MH for replication of the Basic Plant, Byron Nuclear Generating Station. The changes have not been implemented in the installed valves because PSI has not processed the Purchase Order. Sub-sequent to the issuance of this PFR, PSI released Purchase Order Change Notices 36 and 32 for all modified valves (P.O. 1014-88Q for Unit 1 and P.O.
2014-89Q for Unit 2) to incorporate valve upgrade changes.
The results of the physical verification of selected portions of par-tially installed mechanical systems indicate that the mechanical compen-ents, piping, and pipe supports are installed in accordance with the engi-neering requirements.
4-9
)
4.4.2. Verification of Civil Structures The physical verification of civil structures was organized into 69 walkdown packages, involving 208 reference documents. The packages con-sisted of 146 levels of verification (i.e. , 24 level 1, 39 level 2, and 83 level 3). The verification at levels 2 and 3 covered 493 items or rein-forced concrete, structural steel, and cable tray supports. Walkdown pack-ages were prepared by assembling the appropriate engineering design docu-ments necessary to verify construction. The design documents included S&L structural drawings, standards, specifications, Inryco shop / erection draw-ings, Steel Service Company reinforcing steel construction drawings, FCRs, ECRs, and Contractor Drawing Approval Requests. The features to be veri-fied were then color-coded on the design documents. The walkdown package included a set of forms upon which specific items for verification were listed.
After the walkdown was completed, the packages were reviewed for completeness. In some cases, features were photographed as a permanent record.
Table 4-8 summarizes the features selected from the five buildings subjected to the three levels of verification. Twenty-four areas were verified at level 1 for conformance with layout drawings and for identifi-cation of their construction status.
Items verified at levels 2 and 3 were chosen from areas of structures walked down at level 1. Structural items were checked at level 2 for size, location, orientation, and configuration in 39 areas. Structural items were further subjected to detailed dimensional checks, including weld sizes, at level 3 in 83 areas. Table 4-8 illustrates the relationships of these levels of verification and summarizes the number of areas covered in the walkdown. Figures 4-6 through 4-10 show the location of the areas in the selected five buildings.
4-10
( Construction of the selected buildings irriolved four constructors and five material suppliers, as listed in Table 4-9.
Tables 4-10' through 4-12 categorize the types of structures verified.
Table 4-10 lists the types of reinforced concrete and the items covered during the walkdown at levels 2 and 3. Table 4-11' lists the types of structural steel frames and the items covered during the walkdown at levels 2 and 3. Table 4-12 lists the types of cable tray supports and the items covered during the walkdown at levels 2 and 3.
Level 3 verification of reinforced concrete structures included wit-nessing at least one concrete placement at each of the four selected build-ings. The following checks were made during the concrete placement based on Newberg Procedure Manual MPN 9.
e Construction joint preparation before concrete is placed.
e Slope of chutes and use of hoppers for depositing concrete.
e Free-fall distance limited to 3 to 4 ft, and the maximum depth of the horizontal layer deposit limited to 2 ft.
e Step method used when placement covers a large surface area.
e Placement progresses upward from the lower portion on a sloping
( surface.
I
( e Caution used to prevent the concrete from falling directly on the tendon sheathing.
e Time intervals and concrete temperature limited to specification requirements.
{
e Adequate protection provided against inclement weather.
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4-11
{
( - - - - - - - -
I I
e Consolidation of concrete by vibration (e.g. , lateral transporta-tion of concrete with a vibrator is not allowed, vibrator must penetrate previous layer by at least 6 in.).
Level 3 verification of structural steel and cable tray supports included bolt torque strength tests. Bolt sizes ranging in diameter from 7/8 in. to 2-1/2 in. were retorqued at splice locations, connections, and anchor plates.
As a result of the physical verification of selected Unit 1 and 2 structures, four valid PFRs were issued, as follows (see Table 7-1).
PFR 2485:008, classified as an Observation, concerned a minor devia-tion in the location of rebars at the hanger box of the removable roof slab of the Auxiliary Building. These bars were to be located as close as pos-sible to the lifting boxes of the removable slab. However, the rebar loca-tion was not dimensionally specified in the drawing. Subsequently, a Con-tract Drawing Approval Request was issued specifying the rebar location. I This deviatior. did not affect the structural integrity of the removable slab.
PFR 2485:009, classified as an Observation, pointed out an inconsis-tency in connection detail between the contractor's erection drawing and S&L's design drawing. Although shop drawings and erection drawings may contain details different from those shown on the design drawings, these differences are acceptable where the details on member sizes meets or exceeds the requirements of the design documents. The design drawing shows I that stif fener plates are required on a cantilever beam; however, this detail was not constructable and was replaced with an approved alternate shown on the erection drawing. The structural integrity of the connection was not compromised even though the design drawing did not reflect the as-built condition.
l I
4-12
s
(
PFR 2485:010, classified as an Observation, pointed out the construc-
{
tion deficiency in as-built welded studs on embedded plates on the Unit 1 Reactor Containment Building dome liner. Several studs were bent out of
[ tolerance due to, interference with the liner stiffener angles. TPT calcu-lations showed that the stiffeners compensate for the reduction in stud
( capacity, assuming that the bent studs are not effective; therefore, the load-carrying capacity of the embedded plates remains adequate.
[
PFR 2485:013, classified as an Observation, identified out-of-tolerance conditions on rebar spacing and minimum rebar concrete coverage.
This should have been reported in an NCR. This deficiency was found to be an isolated incident and did not affect the structural integrity of the well, since the extent of the out-of-tolerance condition came within S&L's criteria for disposition of NCRs without rework.
The results of the physical verification of selected portions of civil
( structures indicate that the structures are constructed in accordar.ce with the engineering requirements.
[ 4.4.3. QA Support for Walkdown
( This task was designed to provide on-line independent QA surveillance to ensure that walkdown activities were accomplished properly and to pro-vide for immediate corrective action of any deficiencies.
( All 121 walkdown packages were independently reviewed prior to being released for walkdown. This ensured that the packages were complete and that the construction documents included therein were the correct revisions
{
of the latest documents. The post-walkdown QA review of all 121 completed packages verified that the packages had been completed in an accurate and 7 thorough manner. QA reverified 24 walkdown packages, including 17 mechan-ical and 7 structural walkdown packages. The QA reverification consisted of an independent walkdown by QA personnel who were not involved in the original walkdown.
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4-13 T _ - - _ - - - - - - - -
l l
4.5. CONCLUSIONS, TASK C I
Based on the physical verification performed under Task C, the phys-ical installation of selected portions of safety-related mechanical systems and structures of MHNGS Units 1 and 2 conforms to the requirements of design drawings and specifications.
I I
l l
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l l
l l
l 4-14
f L
TABLE 4-1 INSTALLED PIPES PHYSICALLY VERIFIED AT LEVEL 2 FOR UNITS 1 AND 2
- Diameter Diameter Tag No. (in.) Tag No. _ _ _
(in.)
ICC01D18 18 1CC38C6 6 1CC01C16 16 1CC07AB6 6
[ 1CC01B16 16 1CC50F6 6 0CC01A12 12 1CC50E6 6 1CC06A8 8 1CC05C3 3 1CC06DB6 6 1CC05D3 3 1CC02A18 18 OCC02B16 16 1CC02B16 16 1CC50AD3 3 ICC03A16 16 1CC50D6 6
{
1CC02CB12 12 1CC50C6 6 1AF02CH4 4 1CC03FB3 3 1AF02DH4 4 1CC05BB3 3 1AF02ED4- 4 1RY01B6 6 1CC59A16 16 1RY01AA4 4 1CC04B16 16 1RC24AA4 4 0CC04B16 16 1CC03FA3 3 0CC04C12 12 1CC05BA3 3 OCC014D14 14 1CC05G16 16
( 1CC04AB12 12 1CC04AB12 12 1RH01BB12 12 2CC04AB12 12 ,
1RH01CB16 16 2CC02CB12 12 1RH02AB8 8 2RH02AA8 8 1RH10CB3 3 2RH03AA8 8 1RC01A29 29 2RH09AA8 8 1RC01AD29 29 2CC02B16 16 1RC02AD31 31 2CC02CB12 12 27.5
( 1RC03AD27.5 1CC03C8 8
[
4-15
[ I
TABLE 4-2 INSTALLED PIPE SUPPORTS PHYSICALLY g VERIFIED AT LEVELS 2 AND 3 FOR UNITS 1 AND 2 s Level 2 Lebel 3 Level 2 Level 3 Level 2 Level 3 1CC01003R 1CC01003R 1CC22012X 1CC02127X ICC01004X ICC22006X ICC02156R 1CC01001R 1CC13001R 1CC02131X ICC01005R 1CC13003R 1CC02119X 1CC0'119X 1CC01047X ICC13041R 1CC02126R 1CCO2126R 1CC01043R 1CC13016R 1CC13016R ICC02129X ICC02129X 1CC01042X ICC13013R 1RH01018X 1RH01018X 1CC01007R 1CC13042X ICC02010R 1CC02010R 1AB1663A 1CC13051R ICC02010R 1CC02010R 1CC02003R 1CC02003R 1CC13043X ICC02137X ICC02137X 1CC02101X 1CC02101X ICC02004X 1CC02136X ICC03106X ICC03106X ICC02005R 1CC02115R 1CC03023X ICC03057X 1CC03057X ICC02116R 1CC03106X ICC03056R 1CC02167X 1CC03055R 1CC03015X ICC02118R 1CC03003R 1CC03014R 1CC03014R 1CC03006R ICC03016R ISIO6026V 1SIO6026V 1CC03020R 1RH08017X 1RH08017X 1CC03042R 1CC03042R 1RH05009V 1RH05009V ICC03040R 1CC03040R 1RH05005X 1RH05005X ICC03024R 1CC03024R 1CC04028R 1CC03026R 1CC04001X ICC03029R 1CC04036R ICC02163R 1CC04007R 1CC02124R ICC22022X 1CC02125X
- Support type code:
A = anchor S = hydraulic or mechanical seismic ]J C = constant V = variable G = guide W = whip restraint R = rigid X = rigid seismic restraint q 4-16 J
____._____m
[
(
s
- TABLE 4-3
SUMMARY
OF MECHANICAL FEATURES VERIFIED Item Unit 1 Unit 2
[ Pressurizer 1 -
Steam generator 1 -
Heat exchangers 6 1 Pumps 4 1 Motor-operated valves 15 1 Valves (all other) 46 8 Diesel engine 1 1 Electric motor 1 1 Pipe (linear feet) 1380 420 Pipe supports lustalled 44 23 Partially installed 38 12 Not installed 83 6 Inaccessible -
2
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4-17
)
l TABLE 4-4 REACTOR COOLANT SYSTEM EQUIPMENT VERIFIED AT LEVEL 2 FOR UNIT 1 Type Tag No. Supplier Remarks Motor-operated MO-1RC8002D Westinghouse gate valre Motor-operated M0-lRC8001D Westinghouse gate valve Motor-operated M0-lRC8003D Copes Vulcan valve Gate valve IRYO23 Anchor-Darling Relief valve IRY455B Fisher-Control Gate valve IRY022 Anchor-Darling a Sceam generator IRC01BD-lD Westinghouse l Reactor coolant IRC0lPD-lD Westinghouse Only pump housing g pump installed l
Pressurizer 1RY015 Westinghouse l
I I
I I
4-18
(
(
s TABLE 4-5 AUXILIARY FEEDWATER SYSTEM EQUIPMENT VERIFIED AT LEVEL 2 FOR UNITS 1 AND 2 Type Tag No. Supplier Remarks
(
Motor-operated M01AF013H-2 Velan In warehouse globe valve Check valve 1AF014H Anchor-Darling In warehouse Pump 1AF01PB Pacific Pumps Diesel engine 1AF01PB-K GM Pump 1AF01PA Pacific Pumps In warehouse Electric motor 1AF01PA-M Westinghouse In warehouse Air-operated 1AF004B Velan In warehouse globe valve Check valve 1AF003B Anchor-Darling In warehouse Gate valve 1AF002B Velan In warehouse Check valve 1AF001B TRW In warehouse Motor-operated MO-1AF0178-2 Velan In warehouse gate valve Diesel engine 2AF01PB-K GM
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( 4-19 f - - - - - - - - - - - - - - - -- -_
I TABLE 4-6 CCMPONENT COOLING WATER SYSTEM EQUIPMENT VERIFIED AT LEVEL 2 FOR UNITS 1 AND 2 Type Tag No. Supplier Remarks Heat exchanger ICC01A Joseph Oats Gate valve ICC9470B Velan Gate valve ICC9458 Velan Motor- operated M0-1CC9473B-2 Velan gate valve Pump OCC01P Goulds Check valve OCC9464 Velan Gate valve OCC9465 Velan Gate valve ICC9470A Velan Gate valve ICC9467A Velan Gate valve ICC9467B Velan Motor-operated M01CC9415 Velan I gate valve Gate valve ICC9504B Velan Gate valve ICC9502A Velan Not installed Heat exchanger 1FC01A Atlas Industries Gate valve ICC9452C Velan Heat exchanger ICV 04AB-1B Atlas Industry Gate valve ICC9503 Fisher-Control Not installed Butterfly valve ICC9502B velan Not received Butterfly valve ICC9507B Fisher-Control Motor-operated M0-1CC9412B-2 Velan j gate valve l Gate valve ICC9459A Velan Gate valve ICC9459B Velan --
Gate valve OCC9461 Velan On floor of auxiliary building Gate valve 2CC9467B Velan Not installed l Gate valve 2CC9467A Velan Gate valve 2CC9504B Velan -
Butterfly valve 2CC9507B Fisher-Control Not installed 4-20 :
TABLE 4-6 (Continued)
Type Tag No. Supplier Remarks
~
Motor-operated MO-2CC9412B-2 Velan gate valve Gate valve 2CC9459A Velan Gate valve 2CC9459B Velan Air-operated MO-1CC9413B-2 Velan Not installed gate valve Motor-operated MO-1CC9413A-1 Velan Not installed gate valve Globe valve ICC9437A Copes-Vulcan Not installed Check valve ICC9486 Velan Not installed Gate valve ICC9487D Velan
( Gate valve ICC9489D Velan Not installed Globe valve ICC9493D Velan Not installed Gate valve ICC9488D Velan
{
Motor-operated MO-1CC9416-1 Velan Not installed gate valve Motor-operated M0-1CC9414-2 Velan Not installed gate valve Check valve ICC9508 Velan Not installed Air-operated ICC9437B Copes-Vulcan Not installed globe valve Globe valve ICC9511 Velan Not installed Gate valve ICC9452D Velan Not installed Air-operated ICC130B Fisher-Control Not installed butterfly valve Gate valve ICC9499B Velan Check valve ICC9500B Velan Heat exchanger ICV 01AB-1B Atlas Not installed Industries Gate valve ICC9411B Velan Gate valve ICC9499A Velan Check valve ICC9500A Velan
- Heat exchanger ICV 01AA-1A Atlas Not installed Industries Gate valve ICC9411A Velan 4-21 L _ - - - - - -
- TABLE 4-7 RESIDUAL HEAT REMOVAL SYSTEM EQUIPMENT VERIFIED AT LEVEL 2 FOR UNITS 1 AND 2 Type Tag No. Supplier Remarks Motor-operated M0lRH8702A Westinghouse In warehouse gate valve Relief valve IRH8708B Westinghouse In warehouse Centrifugal IRH0lPB2-1B Westinghouse pump Check valve IRH8703B Westinghouse Gate v alve IRH8724B Westinghouse Heat exchanger 1RH02AB Joseph Oats Air-operated 1RH0607 Continental In warehouse valve Motor-operated MolRH8716B Westinghouse Not installed gate valve Air-operated 1RH0619 Continental Not installed butterfly valve Motor-operated M01RH611 Westinghouse globe valve Heat exchanger 2RE02AA Joseph Oats Air-operated 2RH606 Continental butterfly valve Air-operated 2RH618 Continental butterfly valve l
I I
1 4-22 1
w r- v r- --.-_rm w ? ~ m TABLE 4-8
SUMMARY
OF CIVIL / STRUCTURAL VERIFICATION Location an Elevation (ft-in.) Column Line ,1 Level (Quantity) 3 No. Building From To From To From To Description 1_ jL jl 1 Ultimate 425-11 435-0 3 4 D F Concrete slab 1 4 5 Heat Sink and walls 2 Ultimate 435 437-5 1 2 D F Concrete slab 1 1 6 Heat Sink and walls 3 Auxiliary 483 485 11 12 Q S2 Concrete slab 1 1 2 Building 4 Auxiliary 485-6 491-0 6 10 L Q Concrete slab 1 3 5 u Building and walls 5 Auxiliary 492 494-0 26 30 L Q Concrete slab 1 2 3 Building and walls 6 Auxiliary 481 484-6 15 21 Q V Structural steel 1 1 2 Building frame (horizontal frame) 7 Auxiliary 401 417 23 24 L N Structural steel 1 1 3 Building frame (vertical frame) 8 Auxiliary 401 409-6 17 18 Q U Structural steel 1 1 2 Building frame (vertical frame)
TABLE 4-8 (Continued)
Location Plan Elevation (ft-in.) Column Line Level (Quantity)
S ructural No. Building From To From To From To Description 1 2' 3 9 Auxiliary 439 451 11 12 P Q Cable tray 1 1 4 Building support 10 Auxiliary 451 467 15 18 Q U Structural steel 1 1 2 Building (vertical frame) 11 Auxiliary 451 459 12 13 S Q Structural steel 1 1 1 Building (vertical frame) c- 12 Auxiliary 383 --
22 23 P Q Cable tray -- --
I b
Building support (bolt torque testing) 13 Fuel 417 433 21 --
Y BB Concrete wall 1 1 2 Ilandling Building 14 Fuel 417 474 21 --
X BB Structural steel 1( l 4 Ilandling (vertical frame)
Building 15 Containment 579 599 Areas Concrete dome 1 1 2 Building 1 No. I and 4 and ring girder deg segment for level 1 m C ._ m ammuna . m ._
- r- rm.- -.j- v w w M -- -
TABLE 4-8 (Continued)
Location Elevation (ft-in.) Column Line *'* ""'I Y Structural No. Building From To From To From To Description 1 2' 3 16 Containment 579 -
Radius Segment Concrete done. 1 1 2 Building 1 (ft) (deg) (partial) 47 66 270 240 17 Containment 599 --
0 47 270 240 concrete done 1 2 3 Building 1 18 Containment 374 581 Radius Post-tensioned 1 1 2 Building 1 75 -
38,158, (concrete f 278 buttresses)
U 19 Containment 404 --
Plan Structural steel 1 4 4 Building 1 R-12 - - -
(beams and columns) 20 Containment 420 425 R-il R-17 -- --
Cable tray 1 1 3 Building 1 support 21 Containment 440 447 R-ll R --
Cable tray 1 1 5 Building 1 support 22 Containment 377 397-1 Angle (deg) - -
Structural steel 1 2 3 Building 1 180 270 (steam generator support column)
. i
J TABLE 4-8 (Continued)
Location i
Elevation (ft-in.) Column Line Level (Quantity) 3 ,1 No. Building From To From To From To Description 1 2 3 23 Containment 397-1 --
Angle (deg) -- --
Structural steel 1 1 10 Building 1 180 270 (steam generator lateral columns) 24 containmeat 515-5 581-1 90 270 -- --
Concrete 1 3 4 Building 2 (cylinder wall) 25 Containment 401 --
142 --
(Areas 5 Post-tensioned 1 3 3
,, Building 2 and 8) concrete buttress b
(ft-in.)
70, 75-10 Total 24 39 83 (a) Numbers correspond to locatior. numbers in triangles in Figs. 4-6 through 4-10.
(b)These areas are in the same package, r 1 __ r , - amm ammm m
s TABLE 4-9 CONSTRUCTORS AND SUPPLIERS INVOLVED IN CIVIL / STRUCTURAL CONSTRUCTION Constructor / Supplier Name Function Gust K. Newberg Construction Constructor of major structural steel Company and Gust K. Newberg, frame, reinforced concrete structure, Incorporated, Joint Venture and prestressed concrete sheathing /
bearing plates Inryco Supplier of structural steel and prestressed concrete tendon bearing plates Steel Service Company Major supplier of reinforcing steel Chicago Bridge and Iron Supplier and constructor of containment Company liners for walls and domes Westinghouse Electric Supplier of NSSS equipment Corporation Commonwealth-Lord, Joint Constructor of cable tray supports Venture Unistrut Supplier of Unistrut members in cable tray supports 4-27 f --
TABLE 4-10 TYPES OF REINFORCED CONCRETE AND ITEMS COVERED l
4 DURING WALKDOWN AT LEVELS 2 AND 3 Type of Structure Quantity Wall (other than Reactor Containment Building) 11 Column 2 Beam 2 Slab 15 Containment dome 1 Prestressing buttress and ring 7 l Containment wall 1 Total 39 Level 2 Verification Items Major rebar size, spacing, and configuration 34 Tendon sheathing location 8 Rebar protection coverage 16 g Layout plan and elevation 19 Section size 36 Liner plates / stiffeners location, orientation, and 6 configuration Embedment plate location, orientation, and configuration 12 Buttress / ring prestressing bearing plate location and orientation 6 Total 137 Level 3 Verification Items Rebar splice length 24 Stirrup / tie bar details, size, and spacing 8 Construction joint details, size, and location 2 Concrete placement 4 Embedment plate detail 12 Liner plate / stiffener detail 8 Buttress / ring bearing plate detail 5 Total 63 4-28
[
(
TABLE 4-11 TYPES OF STRUCTURAL STEEL FRAME AND ITEMS COVERED DURING WALKDOWN AT LEVELS 2 AND 3 Type of Frame Quantity Column 15 Beam / girder 35 Bracing 8 NSSS equipment support 1 Total 59 Level 2 Verification Items Structural frame overall dimensions 37 Location, orientation, and configuration of steel member 60 Total 97 Level 3 Verification Items Member size 36 Connection detail dimensions 34 Welding size 42 Size and number of bolts 20 Bolt torque strength test 8 Total 140
( l
( 4-29 f ----
i i
- TABLE 4-12 TYPES OF CABLE TRAY SUPPORTS AND ITEMS COVERED DURING WALKDOWN AT LEVELS 2 AND 3 Type of Support Quantity Floor mounted 1 Ceiling mounted 7 Wall mounted 16 Total 24 Level 2 Verification Items Cable tray support identification 21 Location, orientation, and configuration 25 Total 46 Level 3 Verification Items Member size 12 Connection detail dimensions 35 Welding size 23 Bolt torque strength test 12 Total 82 I
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2RH M0 2CC 02AB 2CC 2CC 9507B 1B 9504B 94128-2 RES10UAL 2CC 2CC ICC ICC HEAT-EXCH. FROM COMPONENT 9467A 9467B 9467B 9467A
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- 5. TESTING, TASK D Task D was structured to verify acceptability of ASME pipe welds and concrete in safety-related structures by performing tests and inspections of hardware in conjunction with a review of test procedures and records.
5.1. ASME PIPING WELD INSPECTION, SUBTASK D1 5.1.1. Objective and Scope This subtask verified the acceptability of ASME piping welds by reviewing inspection records and witnessing inspection activities. -
The scope of this subtask included the following:
- Identify and review relevant Marble Hill procedures for weld inspection.
e Identify the specific welds for it.spection.
- Perform visual inspection of selected welds.
e Review radiographic film for selected welds.
- Witness in process radiographic inspection of welds.
A working procedure and checklists were prepared to perform this review and provide an organized means for recording data.
5.1.2. Review of Relevant Marble Hill Procedures All applicable nondestructive examination (NDE) procedures were reviewed for compliance with the requirements of the applicable section and edition of the ASME Code. In addition, procedures were reviewed for 5-1 I -_ - - - - - - _ - - -
evidence of demonstration to, and approval by, the Authorized Nuclear Inspector (ANI). The procedures reviewed were found to be satisfactory.
Table 5-1 lists these procedures.
5.1.3. Weld Inspection Review 5.1.3.1. Selection of Welds. Fifty-four welds were selected from piping segments included in Task C. These included 30 field welds performed by i Cherne and 24 shop welds performed by ITT Grinnell. Table 5-2 lists the welds selected from both Cherne isometrics and ITT Grinnell weld maps. The j welds were selected on available piping in the four mechanical systems verified in Task C. Welds were chosen for the following attributes: q
- 1. Welds that had field radiographic examination by Cherne.
I
- 2. Welds that had shop radiographic examination by ITT Grinnell.
- 3. Different pipe diameters (3, 4, 6, 8, 12, 16, 22, and 24 in.). j
- 4. Different material types (carbon steel and stainless steel).
- 5. Different types of welds (i.e. , pipe to elbow, pipe to valve, I pipe to penetration, T, and flange to penetration).
- 6. Different pipe weld classes (1 and 2).
- 7. Different welds selected from Units 1 and 2.
5.1.3.2. Weld Examination. Fourteen welds were selected for visual examination (see Table 5-3). The 14 welds included those that had ques-tionable surface areas detected during the review of the radiographic film.
Results of the visual weld examination were satisfactory and compared well with the original inspection.
5.1.3.3. Review of Inspection Records and Radiographs. Inspection records and radiographic film (442 pieces of double-loaded film) for the 54 I selected welds were obtained and reviewed, and over 850 individual checks I
I 5-2 l
[
were made. The radiographs were reviewed to the requirements of Cherne and ITT Grinnell procedures and the ASME Code, as applicable. The following attributes were verified:
1 -
- 1. Film density
- 2. Film sensitivity
- 3. Corretc penetrameter selection and placement
- 4. Correct location marker placement
- 5. Adequate permanent weld identification of film
- 6. Correct source utilized per technique sheet
- 7. Geometric parameters per technique sheet
- 8. Weld quality
- 9. Beport for each weld complete and signed off
- 10. Evidence of ANI acceptance PFR 2485:001, classified as a Finding, resulted from this review.
This PFR identified a deficiency on a radiograph of a field weld. Film or source movement occurred during exposure, resulting in a double image. A double image is unacceptable. Regardless of penetrameter sensitivity, film or source movement occurs during exposure, immeasurable geometric factors can occur in the film image, resulting in a loss of weld-flaw-detection sensitivity in the area of interest. The deficiency is attributed to a misunderstanding on the meaning of the film image geometric unsharpness, as defined in the ASME Code.
PSI formulated a CAP stating the following:
( 1. The affected weld will be reradiographed and evaluated for acceptability.
- 2. To prevent recurrence, the applicable procedure will be revised to state that double images will not be accepted on radiographs.
In addition, the inspectors involved will be trained and the training documented.
5-3 f
- 3. The double-image radiograph is considered to be an isolated incident. However, to confirm this, a statistically significant sample of existing radiographs of welds in safety-related systems will b'e reviewed.
Implementation of the CAP will resolve the concern identified in this Finding.
5.1.3.4. In-process Inspection. Radiographic inspections of four welds by Cherne NDE personnel were witnessed (see Table 5-4). Notification for radiography is given by the visual inspector who fills out a Radiographic -
Request /Worksheet Form and tags the weld joint with a white ribbon. The form contains all the necessary information to properly identify the weld I and indicates that the weld has been visually inspected.
The pipe fitter is responsible for permanently identifying the weld by system, weld number, zero location mark, and welder identification. In case of rework, the welder performing the rework also vibroetches his identification with the appropriate rework number.
The following technique parameters were verified: geometric, penetrameter, source position, zone location, proper identification, ade- I quate film size, and permanent identification on weld matched to the Radio-graphic Request /Worksheet.
The in-process inspection indicated that the four welds radiographed by Cherne NDE personnel followed the appropriate procedure and were performed in an acceptable manner.
5.1.4. Conclusions, Subtask D1 One Finding resulted from the review in Subtask D1. A CAP has been _
prepared, and when implemented, will resolve the concern in the Finding, 5
and the procedures and methods used to perform the radiographic and visual 5-4 -
W
[
extminations of ASME welds will be satisfactory and in accordance with the appropriate ASME Code requirements. The selected ASME welds reviewed are adequate.
CONCRETE INSPECTION, SUBTASK D2 f 5.2.
5.2.1. Objective and Scope
{
This subtask inspected concrete in selected safety-related concrete structures using visual inspection and rebound hammer techniques to determine if the structural concrete has any major defects and if the compressive strength of the concrete meets specifications.
( The scope of this subtask included the following:
e Prepare procedures and checklists to facilitate the concrete 3 inspection.
[ e Review cylinder strength data and construction drawings.
e Perform visual inspections and rebound hammer tests (per ASTM C 805) and evaluate the results.
The concrete surfaces of five safety-related concrete structures were inspected visually and tested with a rebound hammer for compressive
{
strength at 50 locations. The inspection covered 25 concrete pours. One of the pours was made in December 1978 before the stoppage of safety-related work, and the remaining 24 pours were made between May 1981 and April 1983, following restart of the safety-related work. Results of cylinder strengths for these pours and the calibration records of the test-ing machine used to test the cylinders were reviewed.
L During this review, more than 100 procedures and design documents were eramined, and over 350 individual checks were made of those documents.
{
( 5-5
( - _ - - - - - - - - --
5.2.2. Procedure Relevant concrete structures were selected for inspection to determine concrete strength using a Soiltest Model CT-320 rebound hammer. The con- I crete pour strength was also verified by reviewing standard cylinder strength tests.
5.2.2.1. Selection of Structures. Five safety-related buildings /
structures, the same structures selected in Task C (Physical Verification-Walkdown), were selected as follows:
- 1. Ultimate Heat Sink for the Essential Service Water Cooling Towers.
- 2. Fuel Handling Building.
- 3. Reactor Containment Building, Unit 1.
- 4. Reactor Containment Building, Unit 2.
- 5. Auxiliary Building.
1 Table 5-5 summarizes the specific areas in these structures which were inspected and tested. The areas were selected to overlap (when practical) with those covered in the walkdown in Task C and to sample various structural members (i.e., walls, columns, slabs) and concrete placement 3
dates. At the time of inspection, all the areas were accessible from existing grade, floors, gratings, or scaffolds.
I 5.2.2.2. Procurement and Calibration of Rebound Hammer. A rebound hammer was procured from Pacific Scientific, Santa Ana, California. The hammer, a Soiltest Model CT-320, S/N 811396, was calibrated at the factory on a Stan-dard Anvil. It was used for tests No. I through 40 at Marble Hill from June 22 to June 24, 1983. On Friday, June 24, 1983, the hammer broke down.
~
The mcat cost.* effective solution was to borrow a similar rebound hammer from Newberg. This was considered acceptable since only 10 more tests had ,
1 5-6 -
i
[-
[ to be performed and the Newberg hammer was of the same brand and model (Soiltest CT-320) as that used by TPT. Newberg hammer S/N 79284 was used in tests No. 41 through 50 on June 27, 1983. Tests No. 11, 12, 17, and 40 were duplicated with the Newberg hammer. The ratio of strengths obtained
[ with the TPT and Newberg hammers ranged from 0.904 to 1.12 and averaged 1.01. This indicated tiat the strengths measured with the two hammers were identical for all practical purposes.
{
5.2.3. Concrete Strength Tests 5.2.3.1. Concrete Cylinder Strength Data. Concrete pour numbers for each location were identified from Newberg drawings No. NMH 10, 16, 17, 19, 21, 28, and 32. The pour dates and pour package numbers were determined from
[ an index card in the Newberg Quality Control (QC) records center. The packages containing cylinder strength data (at age 7, 28 and 91 days) and other QC records associated with the pour were retrieved from the Newberg
{
QC records vault. The package for Pour No. 1CW-EXT-15 had not yet been filed with QC, because the pour was made on April 21, 1983 and the 91-day
{
breaks were not yet performed (due July 21, 1983). Cylinder strength for this pour (at age 7 and 28 days) were obtained from the PSI concrete
[ testing laboratory. All the concrete strength data for the pours identified were reviewed and found to be adequate.
5.2.3.2. Inspection and Test. The concrete surface was examined for
[ overall appearance and major defects at each of the 50 locations. Each area was then tested with the rebound hammer at the rate of 10 readings per test, in accordance with ASTM C805-79, " Standard Test Method for Rebound
{ Hammer of Hardened Concrete." The compressive strength of concrete was determined from these tests.
Table 5-5 lists the building, the test elevation, the pour number, the date of concrete placement, the specified compressive strength of concrete (at age 91 days), the compressive strength of concrete at age 28 and 91
( days as determined by the PSI concrete laboratory on standard 6 by 12 in.
concrete cylinders, i.ne compressive strength of concrete as determined by
[ 5-7 r _ - - - - - - - - - - - --
TPT with the rebound hammer, and the surface condition of the area inspected.
5.2.4. Conclusions, Subtask D2 No Observation or Finding resulted from the review in Subtask D2.
Based on the review performed in Subtask D2, the compressive strength of concrete met specifications in all cases and the areas inspected had no significant defects.
5.3. OVERALL CONCLUSIONS, TASK D Based on the tests and inspections performed and the review of test procedures and records under Task D, the selected ASME pipe welds and selected portions of concrete structures conform to engineering requirements.
The radiographic film deviation detected in this review is considered an isolated procedural deviation which is within the limits of what can normally be expected in any major construction project, and the implementation of the CAP will correct this concern.
I I
I m
5-8 -
r L
r L
I TABLE 5-1 NDE PROCEDURES REVIEWED IN SUBTASK D1
{
Contractor Procedure Title
[ Cherne 14.1.12.7 Nondestructive Examination Personnel Qualification and Certification Program Cherne 14.6 Radiographic Examination of Welds
(
Cherne 14.2 Visible Dye, Solvent Removed Liquid Penetrant Examination Cherne 14.3 Visual Examination ITT Grinnell RTP-1-2A Radiographic Examination Procedure ITT Grinnell QAM 6.1 Nondestructive Examination, Personnel Qualification ITT Grinnell PTP-1-2 Liquid Penetrant Examination ITT Grinnell MTP-1-2 Magnetic Particle Examination ITT Grinnell SIP-92 Nondestructive Examination Instruction Personnel
( Qualification Certification
[
[
[
[
[
[
[
5-9
{
1
[.
TABLE 5-2 WELDS SELECTED FOR SUBTASK D1 Field l System Weld Shop Material Weld Type System Isometric Line No. Spool No. No. Weld Type (a) Welder Procedure of Weld AFS 78401 AFB 004 'lAF02EF-4" MH-05-28X FW4 --
CS NM 1018021-2.25 Pipe to valve AFS 78401 AFB 004 1AF02EB-4" MH-05-28X' --
A CS C532 1-4-2-2 Elbow C520 1-1-1-7 AFS 78401 AFB 004 1AF02EB-4" MH-05-28X --
B CS C150 1-4-2-2 Elbow 1-1-1-7 I AFS 78401 AFB 004 1AF02EB-4" MH-05-28X --
C CS C150 1-4-2-2 Elbow 1-1-1-7 p AFS 78401 AFB 004 1AF02EB-4" MH-05-28X FW3 --
CS NM 101B021-2.25 Pipe to valve g
AFS 78401 AFB 004 1AF02DB-4" MH-05-27X FW7 --
CS NM 1018021-2.25 Pipe to valve AFS 78401 AFB 009 1AF02ED-4" MH-05-66X FW7 --
CS F.D. 101B022-0.56 Pipe to valve AFS 78401 AFB 009 1AF02ED-4" MH-05-66X FW8 --
CS F.D. 1018022-0.56CT Pipe to valve AFS 78401 AFB 011 1AF02DL-4" MH-05-104X FW7 --
CS NP 101B022-0.56 Pipe to valve AFS 78401 AFB 0ll 1AF02DA-4" MH-05-81X --
A CS C204 1-1-1-7 Pipe to 1-4-2-2 pipe AFS 78401 AFB 0ll IAFL2DA-4" MH-05-81X --
B CS C204 1-1-1-7 Pipe to 1-4-2-2 pipe T AFS 78401 AFB 0ll IAF02DA-4" MH-05-81X --
C CS C204 1-1-1-7 Pipe to 1-4-2-2 elbow T j m - o w to o 'o n o u o n i
~ ~ m r- & r- r- . v r- _r- r- .
r- .
rm r-m r- rm 1 m_ rm m TABLE 5-2 (Continued)
Field System Weld Shop Material Weld Type System Isometric' Line No. Spool No. No. Weld. Type (a) Welder Procedure of Weld AFS 78401 AFB 0ll IAF02DA-4" MH-56-81X --
D CS C204 1-1-1-7 Pipe to 1-4-2-2 elbow AFS 78401 AFB 0ll IAF02DA-4" MH-05-81X --
E CS C204 1-1-1-7 Pipe to 1-4-2-2 pipe 1-1-1-7 AFS 78401 AFB 0ll IAF02DA-4" MH-05-81X --
F CS C204 Pipe to 1-4-2-2 elbow AFS 78501 AFB 0ll IAF02DA-4" MH-05-81X --
G CS C204 1-1-1-7 . Pipe to 1-4-2-2 elbow AFS 78501 AFB 0ll IAF02DA-4" MH-05-81X --
H CS C204 1-1-1-7 Pipe to )
y 1-4-2-2 pipe T U RCS 7840lRCA009 1RY01AA-4" MH-10-12X FW6 --
SS MT 8088021-1.0 Pipe to valve ll RCS 7840lRCA009 1RY01AB-4" MH-10-18X FWil --
valve RCS 7840lRCA009 1RY01AB-4" MH-10-18X --
A SS C575 8-4-7-11 Elbow RCS 7840lRCA009 1RY01AB-4" MH-10-18X --
B SS C575 8-4-7-11 Elbow RHR 7840lRHB003 1RH03AB-8" MH-08-46X FW5 --
i RHR 7840lRIIB003 1RH02AB-8" MH-08-43X FW4 --
SS F.S. 808B022-0.56CT Reducer i AFS 7840lRHB003 1RH03AB-8" MH-08-45X FW8 --
SS F.S. 808B022-0.56 Reducer l AFS 7840lRHB005 1RH02AA-8" MH-08-9X FW1 --
SS D.K. 808B021-1.0 Pump l AFS 7840lRHB005 1RH02AA-8" MH-08-11X FW6 --
SS CH, ED 808B022-0.56CT Elbow and HP
l TABLE 5-2 (Continued)
Field System Weld Shop Material Weld Type System Isometric Line No. Spool No. No. Weld Type (a) Welder Procedure of Weld AFS 7840lRilB010 1R1102AB-8" Mil-08-36 FW1 --
pump AFS 7840lRilB010 1R1102AB-8" Mil-08-36 FW3 --
SS E.P. 808B022-0.56CT Valve to valve 7840lRIIB010 1 Ril02AB-8" Mil-08-038 --
A SS C377 8-4-7-11 Elbow AFS C442 C438 AFS 7840lRilB010 1R1102AB-8" Mll-08-038 --
B SS C442 8-4-7-11 Elbow C377 Y 78401CCB313 1CC03FB-3" NY-22-58X FW9 --
CS I.S. 101B021-2.25 Pipe to C v alve CCWS 78401CCB313 1CC03FB-3" NY-22-58X FW8 --
CS 1.S. 101B0212.25 Piie l to valve CCWS 78401CCB315 1CC05BB-3" NY-22-66X FW1 --
CS I.S. 1018021-2.25 Pipe to valve CCWS 78401CCB315 ICC05BB-3" NY-22-66X --
E CS C496 1-4-2-2 Pipe to 1-1-1-7 T CCWS 78401CCB315 ICC05BB-3" NY-22-66X --
F CS C496 1-4-2-2 Pipe to 1-1-1-7 T 78401CVB009 ICV 05B-8" NY-54-54X --
D SS C454 8-4-7-11 Elbow Chem.
Vol. C436 Control 78401CVB009 ICV 05B-8" NY-54-54X -
E SS C454 8-4-7-11 Elbow
- u__; s - - -
- ~ 7- r- r- rm r-m r-m r-m em r- r- v r- r m. rm 11 11 1 --
TABLE 5-2 (Continued)
Field l System Weld Shop Mate ial Weld Type System Isometric Line No. Spool No. No. Weld Type {a) Welder Procedure of Weld Chem. 78401CVB011 1CV13B-3" NY-54-69 FW6 --
SS LV 808B022-0.56CT Pipe to Vol. penetra-Control tion RilR 78402RHB004 2R1101CB-16" ML-08-32X FW5 --
SS I.L. 808B021-1.0CT Elbow Unit 2 78402RilB004 2RH01CB-16" ML-08-32X --
A SS C193 8-4-7-11 Pipe to C312 T 78402RHB004 2RH01CB-16" ML-08-32X --
B SS C300 8-4-7-11 Elbow C438 8-2-7-6 to T
[
w 78402R!lB004 2R1101CB-16" ML-08-32X --
C SS C124 C300 8-4-7-11 8-2-7-6 Pipe to T
C438 78402RHB004 2R1101CB-16" ML-08-32X --
D SS C124 8-4-7-11 Pipe to C300 8-2-7-6 elbow C438 78402RilB004 2R1101CB-16" ML-08-32X --
E SS C193 8-4-7-11 Pipe to C312 elbow 78402RHB009 2R1101BA-12" ML-08-2X FW3 --
SS IL 808B021-1.0CT Elbow 78402RHB009 2R1101BA-12" ML-08-3X FW4 --
SS DK 808B021-1.0CT Elbow Chem. 78402CCdOO2 2CV05B-8" NZ-52-11 F7 --
SS KF 808B022-0.56CT Pipe to Vol. pipe control 78402CVB003 2CV05CA-6" NZ-54-30X FWil --
SS FR 808B022-0.56 Pipe to Unit 2 valve
TABLE 5-2 (Continued)
Field System Weld Shop Material Weld Type System Isometric Line No. Spool No. No. Weld Type (a) Welder Procedure of Weld Chem. 78402CVB003 2CV05CA-6" NZ-54-21X FW12 -- SS FR 8088022-0.56 Elbow Vol. to valve Control Unit 2 78401ECB001 24 in. --
FW1 --
CS LR & 101B021-2.25 Flange to I electrical GP penetra-penetration tion 78401ECB033 24 in. --
FW1 --
CS LQ 101B021-2.25 Flange to '
electrical penetra-penetration tion 7840lPCB008 22 in. --
FW1 --
CS K.K. 101B021-2.51T Flued head to p control sleeve y penetration 7840lPCB016 24 in. --
FW1 --
CS DQ 101B021-2.50 Flued control head to penetration sleeve 7840lPCB010 22 in. --
FW1 --
CS DQ 101B021-2.5IT Flued control head to penetra, tion sleeve (a)CS = carbon steel; SS = stainless steel.
u o u e n o - u o u o o t_; n
- s. .
F L
e L
[
TABLE 5-3
{ WELDS EXAMINED IN SUBTASK D1 Cherne
( Isometric System Drawing No. Weld No. Line No. Spool No.
Chemical 78402CVB003 FW 11 2CV05CA-6" NZ-54-20X and Volume Control 78402CVB003 FW 12 2CV05CA-6" NZ-54-21X AFS 78401 AFB 004 FW 4 1AF02EB-4" MH-06-28X 78401 AFB 004 FW 3 1AF02EB-4" MH-05-28X 78401 AFB 004 FW 7 1AF02DB-4" MH-05-27X 78401 AFB 009 FW 6 1AF02DH-4" MH-08-66X 78401 AFB 009 FW 7 1AF02ED-4" MH-05-66X 78401 AFB 009 FW 8 1AF02ED-4" MH-05-66X
( 78401 AFB 011 FW 4 1AF02DA-4" MH-05-81X 78401 AFB 011 FW 7 1AF02DE-4" MH-05-104X 78401 AFB 004 A 1AF02EB-4" MH-05-28X
(
78401 AFB 004 B 1AF02EB-4" MH-05-28X 78401 AFB 004 C 1AF02EB-4" MH-05-28X Electrical 78401ECB033 FW 1 24-in.-diam penetration
[
{
[
5-15 f - - - - - - -
I I
I l
TABLE 5-4 WELDS SELECTED FOR IN-PROCESS INSPECTION IN SUBTASK D1 Pipe la Weld NDE Size System Line No. Procedure (in.)
CCWS 1CCB313 1 14.6 3 Chemical and Volume Control 2CVB020 4 14.6 3 Safety Injection IS1B011 6 14.6 8 CCWS 1CCB315 8 14.6 3 l
l l
l l
l l
l J
]
5-16
v r-, r- rm rm r- r- r-m r- r- r- rm rm v ,
1 ww TABLE 5-5
SUMMARY
OF CONCRETE TESTS AND INSPECTIONS IN SUBTASK D2 Comprehensive Strength (psi)
Specified Cylinder Test Elevation Pour Pour 91 28 91 Hammer l Structure / System No. Structure (ft) No. Date days days days June 83 Ultimate Heat Sink 1 Wall 371 HSW 369-9 2-16-82 3500 6850 7820 6050 (southwestern 2 Wall 374 HSW 369-9 2-16-82 3500 6850 7820 6700 corner) for 3 Column 372 HSC 369-61 4-20-82 3500 6060 7310 5930 Essential Service 4 Column 372 HSC 369-59 4-20-82 3500 6060 7310 6550 Water Cooling 5 Wall 380 HSW 369-12 4-6-82 3500 6600 7380 5900 9 Wall 372 HSW 369-12 4-6-82 3500 6600 7380 6700 10 Wall 372 HSW 369-12 4-6-82 3500 6600 7380 7400 6 Wall 373 HSW 369-14 6-4-82 3500 4900 7530 6330 7 Wall 372 HSW 369-14 6-4-82 3500 4900 7530 6580 I 8 Wall 372 HSW 369-14 6-4-82 3500 4900 7530 6530 C
Fuel Handling 11 Wall 406 FHW 390-6 3-24-82 5500 6580 7270 6320 Building (fuel 12 Wall 406 FHW 390-6 3-24-82 5500 6580 7270 6400 storage pool) 13 Wall 389 FHW 390-5 11-2-81 5500 6230 7300 6930 14 Wall 389 FHW 390-5 11-2-81 5500 6230 7300 6380 15 Wall 396 FHW 390-5 11-2-81 5500 6230 7300 6830 Reactor Contain- 16 Wall 429 1CW 426-5 12-8-82 5500 6540 7890 6280 ment Building, 17 Wall 429 1CW 426-5 12-8-82 5500 6540 7890 6460 Unit 1 18 Wall 444 1CW 426-5 12-8-82 5500 6540 7890 6100 (inside structures) 19 Wall 456 1CW 441-1 1-13-83 5500 6380 7510 6400 20 Wall top 462 1CW 441-1 1-13-83 5500 6380 7510 5500 21 Wall 381 1CW 377-12 5-20-81 5500 5790 7720 7530 22 Floor 377 1CS 377-1 12-28-78 5500 N/A 6850 7650 23 Wall 382 1CS 377-1 12-28-78 5500 N/A 6850 7180 22 Floor 377 1CS 377-1 12-28-78 5500 N/A 6850 7650 24 Floor 377 1CS 377-1 12-28-78 5500 N/A 6850 6600 25 Floor 377 ICS 377-1 12-28-78 5500 N/A 6850 6750
TABLE 5-5 (Continued) l l
Comprehensive Strength (pei)
Specified Cylinder Test Elevation Pour Pour 91 28 91 llammer Structure / System No. Structure (ft) No. Date days days days June 83 36 Wall 539 1CW-EXT-11 10-30-81 5500 6350 7590 6680 37 Wall 539 ICW-EXT-ll 10-30-81 5500 6350 7590 6930 38 Cutter 597 1CW-EXT-15 4-21-83 5500 6620 --
6300 39 Ring beam 579 1CW-EXT-15 4-21-83 5500 6620 --
6560 Eeactor Containment 26 Wall 381 2CW-377-1 6-11-82 5500 6640 8030 7730 Building Unit 2 27 Wall 381 2CW-377-1 6-11-82 5500 6640 8030 7350 2CW-377-1 6-11-82 5500 6640 8030 7530 (inside structures) 28 Wall 381 7410 30 Wall 381 2CW-377-1 6-11-82 5000 6640 8030 vi 29 Wall 381 2CW-377-6 5-8-82 5500 7600 9700 7030 M 31 Wall 381 2CW-377-2 8-4-82 5500 1900 9260 7320 34 Wall 380 2CW-377-2 8-5-82 5500 7900 9260 7320 32 Floor 377 2CS-377-1 11-13-81 5500 6510 7830 6050 33 Wall 320 2CW-377-12 11-18-82 5500 6320 8460 6480 35 Floor 377 2CS-377-2 10-29-81 5500 6090 7020 6250 Reactor Containment 40 Wall 455 2CW-EXT-6 6-15-82 5500 6410 8250 6060 Building, Unit 2 43 Wall 456 2CW-EXT-6 6-15-82 5500 6410 8250 6750 (outside wall) 41 Wall 418 2CW-EXT-3 12-23-81 5500 6360 8030 6230 42 Wall 418 2CW-EXT-3 12-23-81 5500 6360 8030 6930 Auxiliary 44 Wall 468 AW-451-20 1-3-83 3500 6600 7500 5550 Building (north- 45 Wall 468 AW-451-12 10-27-82 3500 6030 7660 6850 west corner) 46 Wall 468 AW-451-12 10-27-82 3500 6030 7660 6530 47 Wall 455 AW-451-13 6-30-82 3500 7360 8980 7500 48 Wall 456 AW-451-13 6-30-82 3500 7360 8980 7150 49 Wall 443 AW-426-10 6-2-82 3500 6200 8050 7380 50 Wall 443 AW-426-10 6-2-82 3500 6200 8050 6880
_ _ m-__ m m m m
L r
[
- 6. CONSTRUCTION DOCUMENT REVIEW, TASK E This task reviewed ASME piping material certifications, concrete test and inspection records, welder qualification records, and safety-related equipment maintenance and storage records to verify compliance with applic-able specifications and code requirements.
6.1. ASME PIPING MATERIAL CERTIFICATION REVIEW, SUBTASK El 6.1.1. Objective and Scope This subtask reviewed piping and weld filler Certified Material Test Reports (CMTRs) for compliance with the applicable specification requirements.
The scope of this subrask included the following:
e Select documents for review from representative pipe spools.
( e Review selected CMTRs for compliance with code requirements.
A working procedure and checklists were prepared to facilitate the review.
6.1.2. Selection of Certified Material Test Reports Material certifications were selected from 29 pipe spools from piping segments in the following mechanical systems: AFS, RCS, RHR, CCWS, and Chemical and Volume Control. !
6-1
)
ASME Code Class 1 and 2 items were selected for review. Specific attention was given to selecting different size pipelinen, different mate-rial types, and different pipe spools that were part of the ASME piping weld inspection in Subtask D1. Both shop welded (ITT-Grinnell) and field welded (Cherne) pipe stools were selected. Material certifications for three containment penetrations and one electrical penetration were also reviewed.
6.1.3. Review of Material Certifications A total of 193 CMTRs were reviewed for 29 different pipe spools, three
]
containment penetrations, and one electrical penetration. Each CMTR was examined for conformance with the requirements of the ASME, Boiler and Pres- -
sure Vessel Code,Section III, Division 1. The review compared the CMTR with the chemical composition and mechanical properties requirements of the applicable specifications. The CMTR reviews were documented on check-sheets. The comments column noted unusual or nonconforming items.
Table 6-1 lists the system, line number, pipe spool, penetrations, and number of CMTRs examined.
One valid PFR (2485:018), classified as an Observation, resulted from this subtask. The PFR pertained to a weld wire whose weld sample showed 1.28% manganese on the CMTR. The specification limit on manganese content s is 1.0%. PSI will generate an NCR to record the deviation. An evaluation by a TPT metallurgist showed that an additional 0.28% manganese will not adversely affect the structural integrity of the affected welded )
6.1.4. Conclusions, Subtask El Based on the review performed in Subtask El, the selected CMTRs meet ASME Code requirements.
1 6-2
L
[
[
6.2. CONCRETE TEST AND INSPECTION REVIEW, SUBTASK E2 6.2.1. Objective and Scope This subtask reviewed testing and inspection records associated with construction of nuclear safety-related concrete structures for compliance
( with code and construction specification requirements.
The scope of this subtask included the following:
- Select structures for review.
- Identify specifications and procedures.
( e Identify and review records for compliance with codes and specifications.
An evaluation procedure and checklists were prepared for organizing and recording the data.
(
6.2.2. Selection of Structures The testing and inspection records for five safety-related buildings /
structures were selected for review. The five safety-related buildings /
f structures are the same as those covered in Task C (Physical Verification-Walkdown) as follows:
{
- 1. Ultimate Heat Sink for the Essential Service Water-Cooling Towers.
- 2. Fuel Handling Building.
- 3. Reactor Containment Building, Unit 1.
( 4. Reactor Containment Building, Unit 2.
- 5. Auxiliary Building.
(. e-3 L
I I
6.2.3. Specifications and Procedures f l
The following S&L specifications formed the basis of the review:
I
- 1. Y-2721, " Post-tansioning Work," Amendment 9, June 23, 1983.
- 2. Y-2722, " Concrete Structures and Site Work," Amendment 20, April 4, 1983.
- 3. Y-2850, " Inspection and Testing Services," Amendment 16, April 1, j I
1983.
T1-e following procedures, prepared by Newberg, implement the require-ments of the above specifications:
I
- 1. QCP 10.02, " Concrete Preplacement Inspection," Rev. 8, February 16, 1983.
- 2. QCP 10.03, " Concrete Placement Inspection," Rev. 6, March 18, 1983.
- 3. QCP 10.04, " Concrete Curing Inspection," Rev. 4, December 7, 1982 (with Amendments ICN-QCP-10.04-4-1 and -2, February 1, 1983 and February 11, 1983, respectively) .
- 4. QCP 10.05, " Concrete Production Inspection," Rev. 5, March 1, 1983.
- 5. QCP 10.16, "Cadwelding Inspection and Testing," Rev. 2, November 4, 1982.
- 6. QCP 10.17, " Post-Tensioning Embedment Installation Inspection,"
Rev. 1, Juw. 7,1982 (with Amendment ICN-QCP-10.17-1-1, March 11, 1983).
6-4
L F
h
[ Test end Inspection Records Review 6.2.4.
[ Test and inspection records were selected to overlap (when practicable) with areas covered in subtask D2, " Concrete Inspection," and to provide a comprehensive review of inspections and tests associated with construction of safety-related concrete structures at Marble Hill. The review covered
( the following areas of concrete QC.
- 1. Calibration records f'or the testing machine used to test all QC concrete cylinders produced at the site.
- 2. Certification of the batch plant and truck mixers.
- 3. Concrete strength computer plots of individual test and running averages of three consecutive tests.
- 4. Initial qualification tests of portland cement and aggregate.
( 5. Concrete mix design.
- 6. Receiving inspection and certified material test reports for cement, aggregate, reinforcing steel, and bearing plates for s
prestressing tendons.
7.
( Independent control tests on cement, aggregate, and reinforcing steel.
[
- 8. Cadweld inspection, testing, and operator qualifications.
[ 9. Concrete production tests for Pours No. ICW-426-5, ICW-441-1, and 1FW-390-5.
6-5
- _ _ _ . _ _ - __ .1
- 10. In-depth review of inspection and test records associated with Pour No. ICW-EXT-11 (outside wall of Containment Building Unit I at elevation 539 ft), including the following:
- a. Preplacement inspection of rebars, cadwelds, and post-tensioning embednents.
- b. Concrete production, delivery, placement, and curing concrete testing.
v.
- 11. Concrete compressive strength test results for all of the 25 pours investigated in Subtask D2.
Testing and inspection records were reviewed for compliance with code and ,onstruction specification requirements.
In addition, NCRs were reviewed for the 25 pours investigated in Subtask D2. A sample of 37 NCRs (30 from Newberg and 7 from PSI) disposi-
= tioned "Use as is" or " Repair" were selected for review and evaluation.
The selection was made to cover a representative sample of pours, struc-tures, and topics. One NCR was selected for pour No. ICW-EXT-2B (not in Subtask D2) to cover cadwelds. Nonconformances dispositioned " Repair" are sound and those dispositioned "Use as is are for minor deviations that occur commonly in the concrete construction industry and that have no significant effect on the structural integrity of the structures. Table 6-2 summarizes the review of NCRs.
One PFR (2485:012), classified as an Observation, resulted from this subtask. The PFR was concerned with an improperly filed petrog aphic exam-ination test report required for the fine aggregate used in construction of safety-related concrete structures. Subsequent investigations revealed that the petrographic examination had been conducted on the fine aggregate (E.T. Slider Company) used in construction, but was erroneously reported to have been obtained from another company (Scott County Stone Company) in the 6-6
e test report. While the documentation was incorrect, the construction used correctly tested ffne aggregate.
6.2.3. Conclusions, Subtcsk E2 Based on the review performed in Subtask E2, the selected concrete
( testing and inspection records were found to comply with code and con-struction specification requirements.
6.3. WELDER QUALIFICATION RECORDS REVIEW, SUBTASK E3 6.3.1. Objective and Scope
, This subtask reviewed welder qualification records of ASME piping welds for compliance with code requirements.
The scope of this subtask included the following:
[
e Selcet welder qualification records f rom representative welds and welder names.
- Review selected qualification records for compliance with ASME Code requirements.
[ An evaluation procedure and checklists were prepared for organizing and recording the data. During this review, more than 30 procedures and
( documents were examined, and over 150 individual checks were made of these documents.
( 6.3.2. Selection of Welder Qualification Records Welder qualification records were selected for 11 welders (eight from Cherne and three from ITT Grinnell) who welded joints whose radiographs 6-7
l were reviewed in Subtask Dl. Table 6-3 lists the welds, welders selected, and the applicable weld qualification procedure.
6.3.3. Review of Welder Qualification Records Selected Cherne and ITT Grinnell Welder Qualification Records were reviewed to determine the following:
e If the welder is qualified to perform the specified procedure.
- If materials on the Welder Qualification match those actually welded.
o If the metal thickness range to which the welder is qualified falls within that actually welded.
- If the weld is made in a position in which the welder was qualified.
The review was accomplished by using data / checkoff sheets. The records for the welders selected were reviewed for compliance with the ASME Code and Marble Hill specification requirements. Results were compared with ASME Code and PSI specification requirements.
No valid PFR resulted from this review.
6.3.4. Conclusions, Subtask E3 Based on the review performed in Subtask E3, the selected Welder Qual-ification Records meet the requirements of the applicable sections of the ASME Code.
6-8
1 I
L 6.4. SAFETY-RELATED EQUIPMENT MAINTENANCE AND STORAGE REVIEW, SUBTASK E4 6.4.1. Objective and Scope This subtask reviewed the procedures for mainte. nance and storage of safety-related equipment for compliance with requirements.
The scope of this subtask included the following:
- Review PSI procedures for maintenance of installed equipment and for storage of equipment on site.
e Select stored and installed equipment for inspection.
- Review and inspect stored and installed equipment for compliance
( with procedures.
An evaluation procedure and checklist were prepared for organizing and
{
recording data.
~
6.4.2. Review of PSI Procedures Table 6-4 lists the applicable PSI procedures that describe the safety-related equipment maintenance and storage requirements. The applic-able procedures were reviewed to verify that the storage and maintenance program for safety-related equipment complied with the requirements of ANSI N45-2, " Quality Assurance Program Requirements for Nuclear Power Plants,"
{
Section 14, and ANSI N45.2.2, " Packing, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power Plants - During the Construction Phase."
6.4.3. Selection of Stored and Installed Equipment Ten components were selected to be reviewed from those verified in the walkdown of Task C. Five were in storage and five were in place in the s 6-9
plant. Table 6-5 lists the components, applicable Storage and Maintenance Instruction (SMI) number, identification number, and status.
6.4.4. Review and Inspection of Equipment The review used the checklists developed from the applicable PSI SMI for each item of equipment. The requirements in the SMI were compared to the storage and maintenance records for each item of equipment. In addi-tion, each of the 10 equipment items was physically located and its storage status verified. The applicable PSI storage and maintenance procedures were compared to the requirements of ANSI N45.2, Section 14 and ANSI N45.2.2 and were determined to be adequate.
No valid PFR resulted from this review.
6.4.5. Conclusions, Subtask E4 Based on the review in Subtask E4, the applicable PSI storage and maintenance procedures are determined to be adequate and to meet the requirements of ANSI N45.2, Section 14, and ANSI N45.2.2. The review of the storage and maintenance records and inspection of 10 equipment items indicated that they are stored and maintained in accordance with procedures.
6.5. OVERALL CONCLUSIONS, TASK E Based on the review performed in Task E, construction documents selected (ASME piping material certification records, concrete test and inspection records, weldar qualification records, and safety-related equip-ment maintenance and storage records) comply with the requirements of applicable procedures and specifications.
6-10
s E
TABLE 6-1 CERTIFIED MATERIAL TEST REPORTS REVIEWED IN SUBTASK El No.
Identification of CMTRs System Line No. No. Reviewed
, AFS 1AF02EB-4 MH-05-28X 5 1AF02DB-4 MH-05-27X 5 1AF02ED-4 MH-05-66X 5 1AF02DA-4 MH-05-81X 5 1AF02DE-4 MH-05-104X 6 RC3 1RY01AA-4 MH-10-12X 5 1RY01AB-4 MH-10-18X 8 RHR 1RH03AB-3 MH-08-46X 6 1RH02AB-8 MH-08-43X 7 1RH03AB-8 MH-08-45X 7 1RH02AA-8 MH-08-9X 7 1RH02AA-8 MH-08-11X 5 1RH02AB-8 MH-08-42X 7 1RH02AB-8 MH-08-36X 5 1RH02AB-8 MH-08-38X 5 2RH01BB-12 ML-08-31X 6 2RH01CB-16 ML-08-32X 8 s
2RH01CB-16 ML-08-35X 8 2RH01BA-12 ML-08-2X 7 2RH01BA-12 ML-08-3X 5
{
CCWS 1CC03FB-3 NY-22-58X 3 1CC03FB-3 NY-22-57X 3 1CC03FB-3 NY-22-66X 6 Chemical and Volume Control 1CVA1A-6 NY-54-55X 4 1CV05B-8 NY-54-54X 8 6-11
TABLE 6-1 (Continued)
No.
Identification of CMTRs System Line No. No. Reviewed Chemical and Volume Control 1CV13B-3 NY-54-69X 8 (continued) 2CV05B-8 NZ-54-11X 6 2CV05CA-6 NZ-54-20X 6 2CV05CA-6 NZ-54-21X 6 Penetration Chilled water 1W004DB-10 1PC10 5 190140B-10 1PC8 5 Containment spray ICS02AB-10 1PC16 6 Electrical IRC02E 5 6-12
n H
TABLE 6-2
SUMMARY
OF NONCONFORMANCE REPORTS REVIEWED IN SUBTASK E2 Structure Total Reviewed System (a) Item Pour No. NCRs NCR No. Date Description
) UHS 8 HSW-369-9 5+1 4806 10/14/82 Concrete surface j scratched
. 11 HSW-369-9 --
182-QIT-0042 02/18/82 Cylinders moved too soon after casting 10 HSW-369-12 3+0 3246 03/12/82 Rebar spacing out of tolerance 13 HSW-369-12 --
3440 04/08/82 Curing temperature not recorded 15 H3W-369-12 --
3428 04/07/82 Air content too low 9 HSW-369-14 2+1 3515 04/15/82 Not enough concrete on rebar 12 HSW-369-14 --
182-QIT-0082 05/10/82 Cylinder curing too hot 16 HSW-369-14 --
3250 03/15/82 Dowels bent out of tolerance 14 HSW-369-59 2+0 3583 04/23/82 Air content too low HSW-369-61 2+0 None --
RHB 1 FHW-390-5 11+2 2315 11/04/81 Concrete tempera-ture high by 2*F j 1 FHW-390-6 4+0 5677 03/15/83 Wall thickness out of tolerance 3 FHW-390-6 --
3072 02/19/82 Rebar spacing out of tolerance s RCB --
ICS-377-1 9+0 None --
nside 32 1CW-377-12 6+0 4062 06/26/82 Mislocated embedded p y ,t, 28 1CW-441-1 4+0 5435 01/12/83 Joint filler installed too early 33 1CW-441-1 --
4929 10/29/82 Mislocated studs 29 1CW 426-5 6+3 5176 12/02/82 Too much concrete cover on rebar 30 ICW 426-5 --
5204 12/06/82 Unknown lubrication on steam generator embedment 6-13
TABLE 6-2 (Continued)
Structure Total Reviewed System (a) Item Pour No. NCRs NCR No. Date Description 31 1CW 426-5 -- 6121 06/21/83 Not approved procedure for lube 36 1CW 426-5 --
182-QIT-0180 12/30/82 Lost thermometer RCB 35 1CW-EXT-11 3+1 181-QIT-0058 11/03/81 Cylinders cured too Unit I hot Outside 37 1 CW-EXT-11 3416 04/05/82 Tendon tube kinked and out of position 34 1CW-EXT-15 6+0 5908 04/29/83 Nicked rebars and anchors 27 1CW-EXT-2B - 4513 08/24/82 Wrong bars cadwelded RCB 26 2CW-377-1 8+1 4261 07/20/82 Rebars misplaced to Unit 2 miss sleeve nside 4933 11/01/82 Embedded plate out 20 2CW-377-2 11+2 of tolerance 23 2CW-377-2 -- 5595 02/ /83 Embedded plate out of tolerance 21 2CW-377-6 0+1 182-QIT-0130 08/06/82 Cylinder curing too cold
-- 2CW-377-12 0+2 None -- --
17 2CW-377-1 10+1 16-22-S 07/15/81 Inadequate splice NCRs 18 2CS-377-1 --
1736-S 08/03/81 No rebar traceability 22 2CS-377-1 -- 182-QIT-0036 02/15/82 Cyclinder curing too hot 24 2CS-377-2 10-1 2302 11/03/81 concrete slump too high 17 2CS-377-2 -- 1622-S 07/15/81 Inadequate splice RCB 25 2CW-EXT-3 2+1 2530 12/15/81 Excessive concrete
> Unit 2 cover on rebar 19 2CW-EXT-6 6+3 5161 12/01/82 Tendon tube partially obstructed AB --
AW-426-10 0+0 None -- --
4 AW-426-12 4+1 4777 10/12/82 Rebar bond broken 6-14
1 L
r b
TABLE 6-2 (Continued)
Structure Total Reviewed System (a) Item Pour No. NCRs NCR No. Date Description 5 AW-426-12 -
182-QIT-0160 11/15/82 Air meter out of calibration 6 AW-426-12 -
5691 03/18/83 One NCR not closed before pouring AW-426-13 0+0 None - -
7 AW-426-20 1+0 5282 12/17/82 Dowels missing
- URS = Ultimate Heat Sink for Essential Service Water cooling towers; FHB = Fuel Handling Building Fuel storage pool; RCB = Reactor Containment Building; NCR = Nonconformance Reports (summary);
, AB = Auxiliary Building, northeast corner.
I
{
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TABLE 6-3 WELDER QUALIFICATION LIST FOR SUBTASX E3 Welder Identifi-Weld cation Employer Weld Procedures Line No. Spool No. System FW5 IL Cherne 808B021-1.0(CT) 2RH01CB-16" ML-08-32X RHR FW9 LS Cherne 101B021.25 1CC03FB-3" NY-22-58X CCWS FW4 NM Cherne 101B022-2.25 1AF02EB-4" MH-05-28X AFS FW5 MT Cherne 808B021-1.0(CT) 1RC01AA-4" MH-10-12X RCS FW6 CH Cherne 808B022 .56(CT) 1RH02AA-8" MH-08-11X RHR FW6 HP Cherne 808B022 .56(CT) 1RH02AA-8" MH-08-11X RHR FW7 FD Cherne 101B022 .56(CT) 1AF02ED-4" MH-05-66X AFS FW6 ED Cherne 808B022 .56(CT) 1RH02A-08" MH-08-11X RHR SWA C-532 ITT Grinnell 1-4-2-2/1-1-1-7 1AF02EB-4" MH-05-28X AFS SWB C-150 ITT Grinnell 1-4-2-2/1-1-1-7 1AF02EB-4" MH-05-28X AFS SWA C-520 ITT Grinnell 1-4-2-2/1-1-1-7 1AF02EB-4" MH-05-28X AFS s
6-16
)
t L
TABLE 6-4 PSI PROCEDURES FOR SAFETY-RELATED EQUIPMENT MAINTENANCE AND STORAGE REVIEWED IN SU3 TASK E4 Effective No. Title Date 7-5 Receipt Inspection 03/30/83 7-6 Release of Material and Items 05/12/83 7-9 Conditional Release of Material and Item 03/30/83 7-10 Receiving of Material and Itema 01/17/83 7-11 Storage and Maintenance Surveillance 11/09/82 7-12 Materials Management Information System 03/30/83 7-19 Storage Maintenance Information System 05/25/83 13-1 Storage, Handling, and Maintenance of Materials and 06/03/83 Items 13-2 Storage and Warehousing 03/30/83 14-1 Inspection Status 10/26/82 17-3 Procurement Data Packages 06/17/83 6-17
TABLE 6-5 EQUIPMENT SELECTED FOR STORAGE AND MAINTENANCE RECORDS VERIFICATION IN SUBTASK E4 Storage and Maintenance Identification Component Instruction No. No. Status Check valve Y-2718C-001 1AF003B Stored Check valve Y-2718B-001 1AF001B Stored Gate valve Y-2718A-001 1AF002B Stored Air-operated valve Y-2718A-001 1AF004B Stored Motor-operated valve Y-2718A-001 1AF0017B Stored Check valve Y-2718C-001 1AF014H In place Gate valve Y-2702-006 1CC9467A In place Gate valve Y-2702-006 1CC9504B In place Feedwater pumps Y-2758C-001 1AF01PB In place Feedwater pumps Y-2758C-001 2AFG1PB In place 6-18
\
- 7. OVERALL
SUMMARY
AND CONCLUSIONS TPT conducted an independent review of the QA organization and construction of MHNGS. The review was organized into several tasks. Task A evaluated the adequacy of the PSI QA organization and the effectiveness of management policies in promoting quality consciousness in the construc-tion of Marble Hill. The major program effort evaluated the adequacy of the construction process in converting the engineering design into the con-structed hardware. Task B reviewed the construction deign control system and its implementation. Tasks C and D verified the physical installation of selected portions of safety-related mechanical systems and structures.
Task E reviewed selected construction documents.
7.1.
SUMMARY
OF OBSERVATIONS AND FINDINGS This program used PFRs to document and resolve geestions raised during the review process. This mechanism was highly formalized to ensure that no pressure could sway the reviewer's technical judgment, allowing the reviewer to raise any potential concern. PFRs were reviewed to ensure that questions raised were accurately communicated and that pertinent inforna-tion had not been overlooked by the reviewer. Still, some PFRs were f.niti-ated because reviewers lacked information or had an inadequate understand-ing of the process or approach used by PSI, S&L, Westinghouse, Cherne, or Newberg in the area of concern. Thus, 8 of the 19 PFRs were satisfactorily answered during the process and were declared Invalid. Table 7-1 summarizes all 19 PFRs resulting from this program.
i Of the 11 valid PFRs, nine were Observations and two were Findings.
The nine Observations were distributed between Tasks B, C, and E (1, 6, and 2 respectfully). The two Findings were associated with Tasks A and D.
7-1 l
The 11 Observations and Findings f all into four types of deviations:
(1) inadequate procedure, (2) procedure not followed, (3) installation not to engineering documents, and (4) improper documentation.
Six Observations (PFR 2485:007, 008, 010, 013, 016, and 018) were of type 3 (installation not to engineering documents). In all cases, the impact of the difference between the installed condition and the design was accommodated within design margins. In one case, engineering drawings had been changed, but the installation had not yet reflected the changes, because the Purchase Order for the hardware had not been processed.
One Observation (PFR 2485:005) was of type 1 (inadequate procedure),
concerning a vagueness in PSI project procedures regarding the delegation of certain design review functions. Clarification of the procedure will correct this deviation.
Two Observations (PFR 2485:009 and 012) were of type 4 (improper documentation). In both cases, either an engineering document or report had not been updated to reflect an approved or acceptable condition in the field.
The two Findings (PFR 2485:001 and 011) resulting from this review are ot type 2 (procedure not followed). One Finding (PFR 2485:001) questioned the acceptance of a weld inspection radiograph containing a double image.
If the x-ray source or the film is moved during exposure, geometric imper-fections occur in the radiograph which could result in a loss of flaw detection sensitivity in the weld area under inspection. Industry practice is to reject all double image radiographs and reradiograph the weld to eliminate the possibility of obscuring flaws in the weld.
The second Finding (PFR 2485:011) addressed concerns regarding PSI management audits of the QA program. Review of management audit reports over the last three years showed that some of the audit reports were limited to an evaluation of the implementation of the QA program and did 7-2
not, in the opinion of the TPT reviewer, contain a substantive statement about the effectiveness of the program. Also, one audit did not address
- compliance with applicable codas, standards, and Regulatory Guides.
The following section discusses CAPS prepared by PSI addressing the concerns identified in the two Findings.
f 7.2.
SUMMARY
OF CORRECTIVE ACTION PLANS PSI prepared a CAP for each of the two Findings resulting from this conttruction review. These CAPS describe the planned approach to correct deviacions identified in the Findings. TPT reviewed the CAPS to ensure that the deviations were properly understood, that the implemented CAP would remove any concern identified in the Finding, and that it addressed possible generically similar deviations.
The CAP submitted by PSI for the Finding (PFR 2485:001) concerning weld inspection radiographs will require the affected weld to be reradio-graphed. Recurrence will be prevented by revising the radiograph inspec- .-
{ ,
tion procedure to make double-image radiographs unacceptable, training the inspectors accordingly, and performing a reexamination of existing radio-graphs to determine if any other double images exist using a statistically rigorous sampling plan.
The CAP addressing the concerns in the Finding (PFR 2485:011) on management audita of the QA program requires a change (which has been effected) in the Project Management Procedures to ensure that a more sub- '
stantive statement regarding QA program effectiveness is made and that future audits check compliance of the QA program with codes, standards, and Regulatory Guides.
1 Both CAPS demonstrated that the deviations in the Findings were indeed i properly understood, and when implemented, the planned action, taken in concert with the rest of the program, would remove the concerns that the Findings may - have raised.
7-3
(
In summary, both Findings will be satisfactorily closed out upon implementation of the CAPS.
7.3. CONCLUSIONS This program for an independent construction review of MHNGS covered a broad range of activities of PSI, the managing / operating agent and con-struction manager, and its major civil / structural and mechanical contrac-tors. The various aspects of QA and construction process reviewed were representative ot activities since resumption of safety-related work at Marble Hill in 1980. At the time of the review, construction was approxi-mately 50% complete on Unit 1 and 25% complete on Unit 2. The conclusions on status of the QA organization, the attitude of PSI management toward quality, and the construction change control system encompass the entire Marble Hill Proj2ct, while the conclusions on the construction of hardware are specific to the type of hardware actually verified.
The construction review of Marble Hill from various perspective provides a basis for the following conclusious:
- 1. The status of the PSI QA organization is considered to be at a level consistent with that required for an ef fective surveillance of QA activities at Marble Hill. PSI management policies affect-ing quality, augmented by extensive involvement and interest by management in QA activities, have a positive influence in ensuring quality in the construction of Marble Hill.
- 2. The PSI system for controlling design changes during construction is adequate, and the procedures were rigorously implemented in the construction of the selected mechanical systems and structures reviewed.
Other contractor activities were not reviewed (e.g., electrical installation by Commonwealth-Lord, Joint Venture) due to the very early stage of construction.
7-4
) f~
l L
- 3. Construction of the selected portions of safety-related mechanical systems and structures conforms to engineering j requirements, based on on-site physical verification of hardware, verification by tests and inspections, and review of construction documents. .
Overall, the project segments reviewed by TPT are indicative of a project with good management attitude and policies toward quality and an adequate QA organization, resulting in adequate construction at MHNGS.
e 1
7-5
?
TABLg 7-1 FOTENTIAL FINDING REFGET SUMMART Task / Responsible Systes/
Description of Concern Disposition of Concern Classification PFt No. Subt ask Organisation Structure Field pipe weld inspection A CAF has been formulated to Finding 2445:001 Di Cherne RER radiograph showed that film (1) reinspect the weld. (2) or source movement had occur- reinspect a statistically red during esposure, result- significant sample of all ing ta a double image en the pipe weld radiographs in film. Inountry practice saf ety systems to ensure would have required that no more double-taage reinspection. radiographs have been accepted. (3) revise proce-dures so no double-image radiographs will be accepted in the future.
ITT Grinnett AFS Zone location markers on a The tone location marker did Inva: 14 2485:002 01 shop wela radiograph were not aaversely af f ect viewing located in the area of area; it to acceptable. Also.
Interest. this is an isolated incident.
ITT GrinneLL AFS The Marble Hill site ha* in- Further investitFlon at ITT Invalid 2485:003 E3 sufficient documentatic Crinnell's welding shop showed welder qualification ts that there was adequate defini-determine if ITT Grinneli .tten of weld procedures for welders were quellfied to sch step of the weld, and weld-specific welding procedures. ers were qualified to these procedures.
2485:004 B PSI N/A Procedures defining external PSI defines these interf ace Invalid interf ace activities appeared requireeents in the contracts to be f ragmentary. and procurement documents with individual contractors.
PSI N/A FSI has delegated design PSI is going to change the Observatica 24811005 S review responsibility to S&L; wording in the FQAM to pr9-however. FS! has retained vert this aisunderstanding in certain design review the future.
functions according to their FQAM. The pries design review responsibility is unclear.
2485:006 A FSI N/A Job / position description for A review of the qualifications Invalid PSI QA management personnel of the current QA esnagers appeared to be deficient in shows thee to be edequate.
identifying the minimus PSI has new job descriptions qualifications. La process to be released in the near future.
2485:007 C Cherne CCWS The as-built location for The condition will require an Observation pipe restraist was inconsis- NCE against the support.
tant with the specification. Strees analysis eerformed by TFT shows acceptable conditions.
Newberg Aum111 sty Robers were aislocated at the The design drawieg was not Obse rvation 2485t006 C Building hanger boa of the roof remov- updated; however, structural able slab. integrity is not af fected.
2485:009 C S&L Aux 111ary Connection details in design The as-built condition was Observation Sutiding drawings were inconsistent based on the shop drawings with shop drawings. which had been reviewed and approved by S&L. However, the design drawing does not reflect the as-built condition.
CBI RCS Unit i As-built welded stude on Con = Stress analysis showed that Observ ation 2 A 5: 010 C tainment Building does liner the structural capacity was embed plates were bent out of not degraded.
tolerance.
7-6
/
l TABLE 7-1 (continued) 1 Taek/ Responsible Systee/
- FFE No. Subt ask Organisation Structere Description of Concern Disposition of Concern Classification 2485:011 A PSI N/A Although management audits A CAP was instituted by PSI Finding have been performed annually that ensures that future as required, a review of soes management audits will require reporta indicated that audits more substantive staten=nts
(
I were limited to evaluating regarding program ef fectiveness the implementation of the QA and that project sanagement program and did not contain a procedures have been endtfied substantive statement on the to assure that audits include ef fectiveness of the program. review of PQAM's for compliance One audit did not address with applicable codes, stan-compliance with applicable dards and Regulatory Guides.
codes, standards, and Regu-latory Guides.
2485:012 E2 Newberg Category 1 N petrographic examittation Paperwork covering fine aggre- Observation concrete for fine aggregate used in gate testing was in error.
structures Category 1 concrete was not Petrographic esaatnatise was verifiable. conducted on the sand used in conet ruction.
2485:013 C Newberg FHB The reber spacing and minimas The structural integrity le Observation concrete coversgo escoeded the not af fected and this appears tolerance limit. to be an isolated incident.
2485:014 C Newberg RCS As-euilt square tendon anchor The anchor plate in question Invalid Unita 1 plates were 1-in. short in was for vert 1eal, rather than and 2 both directions. horisontal, tendons and was the correct else.
2485:015 C Westinghouse RHR The motor on the actor- Correct rpe to 1700 and the lovalid operated valve was 1700 rpe drawing discrepancy was when the drawing called for already noted by project cor- ,
3600 rpe. respondence P5W-2757. The correct valve motor is installed.
2485:016 C PSI CCWS Twelve valves were upgrs&:d PSI has released the Purchase Observation by Westinghouse. Related Order for the valve upgrade.
hardware had not been installed.
f 2485 017 to PSI AFS No documentation could be The required documentation Invalid located te show that the existed at the time of the quarterly maintenance on review but was not located feedwater pump 2AF01Pt had untti af ter this FF1 was in been performed as required. process.
2485:018 11 Cherne N/A ASME SFA 5.23Al limits An NCE will be generated and Observation manganese content to 1.01. processed by PS1 on this devi-The sample shows the mangen- ation. However, evaluation ese content to be 1.282. of the ef fe.& of an additional ,
0.281 Ma by metallurgist show no adverse af fect.
24P5:019 El Cherne N/A A material certification for two dif ferent veld wire heat Material certification wme located.
Invalid QA numbers could not be located.
(a)RCS = Reactor Containment Building; FRB = Fuel Mandling Building. RK1 = Reeldual Heat Removal Systee; AFS = Auxiliary Feedwater Systee; CCWS = Component Cooling Water System.
7-7
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _