IR 05000424/1986027
| ML20215E846 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/08/1986 |
| From: | Blake J, Novak T, Sinkule M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215E832 | List: |
| References | |
| 50-424-86-27, NUDOCS 8612230143 | |
| Download: ML20215E846 (48) | |
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r 4 ;/ UNITE 3 STATES m H:oq'o NUCLEAR REGULATORY COMMISSION y* , REGION ll
_g j 101 MARIETTA STREET, * 2 ATLANTA, GEORGI A 30323
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Report No.: 50-424/86-27 o Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302
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Docket No.: 50-424 Construction Permit No.: CPPR-103 Facility Name: Vogtle 1 Module: No. 16, Nuclear Steam Supply System Reviews Conducted: At various times from March 25 - August 8, 1986
.0n-Site Inspections Conducted: May 5-9, 1986; June 2-6, 15-19, 22-25, 1986; July 7-11, 20-24, 28-31, 1986 NRC Offices Participation in Inspections / Reviews:
Region II, Atlanta, GA
.0ffice of Nuclear Reactor Regulation (NRR), Bethesda, M0 Office of Inspection and Enforcement (IE), Bethesda, MD Inspectors: -John W. York, Senio'r Resident Inspector, Region II Steven J., Vias, Reactor Inspector, Region II John Menning, Reactor Inspector, Region II Reviewer: so , NRR Approved b ,_ , 9 E6 J. J BTake, Chief (Module - Section 6.1) Date Signed .
M t rials and Process Section i ision of Reactor Safety, Region II ,
^O u (Lh.' l2- 8 AC V., SLnkdle, Chief - (Module - All Sections) Date/SiQned Projects Section 20 Division of Re or Projects, Regfon II
_ -L L L n > ** b Ob T. Novak, Deputy Director
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Date/ Signed Division of PWR Licensing A Office of Nuclear Reactor Regulq on (NRR)
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CONTENTS Topic Page Summary iii Scope of Review 1 Methodology 1 Evaluations 3 Findings 37 Conclusions 38 Tables 42
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V0GTLE ELECTRIC GENERATING PLANT UNIT 1 READINESS REVIEW PROGRAM MODULE 16 NUCLEAR STEAM SUPPLY SYSTEMS SUMMARY Georgia Power Company (GPC) is conducting the Readiness Review Program to assure that all design, construction, and operational commitments have been properly identified and implemented at the Vogtle Electric Generating Plant (VEGP) Unit Module 16, which was submitted on January 9,1986, presents an assessment of compliance with licensing commitments for the nuclear steam supply system. The evaluation described in this report was conducted to determine whether the results of the Readiness Review Program review of the nuclear steam supply system presented in Module 16 represent an accurate assessment of the requirements, that the requirements were properly implemented, and that the resolutions to findings were appropriat The evaluation was performed by NRC inspectors from Region II, along with reviewers from the Office of Nuclear Reactor Regulations (NRR). An additional evaluation of the Independent Design Review (IDR) will be performed by the Office of Inspection and Enforcement (IE). The NRC Region II and NRR evaluation was accomplished through a detailed examination of all sections of Module 16 by: Verifying that commitments related to the nuclear steam supply system were identified in accordance with the FSAR and other commitment sources, Verifying that a sample of these commitments were implemented in design and construction procedures, specifications, and design criteria, Checking a representative sample of the documents reviewed by Readiness Review personnel and an independent sample of documents selected by the NRC inspectors, Checking a representative sample of nuclear steam supply system components installed in Unit 1, and supporting quality control (QC) documentation, Reviewing results from previous NRC inspections at Vogtle Unit 1 pertaining to Module 16, Reviewing Module 16 findings and their resolution During this evaluation it was apparent to the NRC Region II inspectors that GPC management supported the Readiness Review program by active participation in the development and implementation of the program. This evaluation also indicates that the applicant's program review was comprehensive and provides adequate assurance that, except for several findings identified by the NRC Region II iii t
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inspectors, nuclear steam supply system for the Vogtle plant is in accordance with NRC requirements and FSAR commitment The findings are subject to continuing review until proper resolution has been reache The findings identified during the evaluation are summarized in the three items listed below:
o Deficiency - Incorrect main steam generator nozzle load allowables and reference specification, (VIOL 424/86-78-01).
o Deficiency - Failure to follow procedure for documenting deviations, (VIOL 424/86-78-02)
o Deficiency - Inadequate review of calculation for steam generator main steam nozzle loads, (DI 424/86-78-03).
The foregoing do not appear to represent significant programmatic weaknesse This conclusion is reached under the condition that these open items for VEGP Unit I can be satisfactorily close Resolutions to all open items will be handled during future NRC Region II inspection iv
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s r V0GTLE ELECTRIC GENERATING PLANT UNIT 1 READINESS REVIEW PROGRAM MODULE 16 NUCLEAR STEAM SUPPLY SYSTEM Scope of Review This review consisted of an examination of each section of the module and was performed by inspectors from Region II along with reviewers from the Office of Nuclear Reactor Regulation (NRR). Module 16, Sections 1.0, 2.0, 4.0 and 8.0, presented data on module organization, project organization, program description, and conclusions. These did not require as extensive a review as that given Module 16, Sections 3.0, 5.0, and 6.0, which covered commitments, audits and special investigations, and program verificatio The latter three sections contain the more significant aspects of the program relating to licensing commitment These include commitment identification, commitment implementation into the design and construction programs, and technical adequacy of the design in meeting the commitment The evaluation of these three sections included an examination of content, review of findings and conclusions, review of a sample of items previously reviewed in the Readiness Review Program and review of an independently selected sample of field construction. The methodology used and an evaluation of each section are presented in the following.
( Methodology Region II Review The NRC review performed by the Region II inspectors was concerned with all sections of the report with the exception of Section 7, but focused on Section 5 and Subsections 3.5, 6.1, and The entire Module was read and. reviewed for organization and content after its receipt in January 1986. This was followed by seven inspection trips to the site on May 5-9,1986; June 2-6,15-19, 22-25,1986; July 7-11, 20-24, and 28-31, 198 General review activities included an assessment of the information presented in Section 1, 2, o and 8 of Module 16. Questions concerning these sections of the module were discussed with the Readiness Review Team personne Section 3 is divided into subsection Subsection 3.4, Commitment Matrix, consisted of 95 commitments relating to the nuclear steam suppy system. The review of this subsection was performed by NRR. The NRC Region II inspectors reviewed all of the correspondence concerning the NRR review of thir subsection and the resolution of any problem areas i
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raised. For subsection 3.5, Implementation Matrix, the Region II inspectors reviewed 37 commitments of the 95 listed for verification of commitment implementation from commitment source documents to first order implementation document Section 5, Audits and Special Investigations, was reviewed for general-content since this section contains a description of the audit program, INPO evaluations, and NRC inspections. The NRC Region II inspectors reviewed the composite list compiled by the Readiness Review Team of the most significant findings resulting from the audits, evaluations and inspection In addition, a review of some audits in their entirety was performed in order to assure that the most important findings had been included in the Readiness Review Teams lis The Design Program Verification, Subsection 6.1, concentrated on the design interface (i.e. the flow of design information) between Bechtel Power Corporation and Westinghouse. An evaluation of the Design Program Verification was made by reviewing a portion of seven areas out of eleven areas total examined by the Readiness Review Team. The NRC Region II inspectors examined the phase II part of the seven areas. In phase II design documents were reviewed to ascertain whether required design interface data had been properly transmitted, received and implemente An evaluation of the Construction Program Verification, Subsection 6.2, dealt with nuclear steam supply system installation and related construction activities performed by Nuclear Installation Services Company (NISCO). The NRC Region II inspectors performed a field inspection for installation activities that had occurred on one of the four steam generators, the pressurizer, and the bottom mounted instrumentation guide tubes. In addition to this field inspection, a review of the documentation, procedures, materials control, personnel certifications (both welders and QC personnel), nondestructive examination, and nonconformance handling were reviewed for these components, NRR Review The NRR review of the Vogtle Readiness Review Module 16, Nuclear Steam Supply System was initiated by a review of the commitment matrix that was conducted in three phases. Phase I of the review consisted of a comparison of the commitments listed in Section 3.4 of the module and with the guidance contained in the Standard Review Plan Sections 1.9, 3.1, 5.2, 6.1, 6.2, 6.6, 9.2, and This phase of the review ensured that all Module 16 commitments accurately represent the status of the licensing requirements. The staff also reviewed the commitments to ensure that all commitments necessitated by the Standard Review Plan were included in the Readiness Review Program (not necessarily in this module however). Phase II of the review consisted of discussion of the commitments as compared with the source FSAR section listed for each
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commitment. As a result of this phase of review, several omissions and apparent errors were identifie These items were discussed _for clarification and then forwarded to the applicant in the form of staff questions to be answered. Phase III consisted of an evaluation of the applicant's response to questions submitted in Phase II. The applicant satisfactorily responded to these questions and either modified the commitment matrix or revised the FSAR to reflect the correct Standard Review Plan positio IE Review The review to be performed by IE will focus on the Independent Design Review (IDR) reported in Section 7.0 of the Module and will be completed and addressed in Readiness Review Module 22, Independent Design Review. Section 7.0 will be read for general content, coherence, and completeness. The IDR findings and corrective c.ctions will be reviewed to verify validity and implementatio Specific design calculations will be selected and independently reviewed for technical adequacy. Selected construction drawings will also be examined for correct transfer of design information to the drawing . Evaluation The evaluation of each Module section is provided below in the section-by-section format used in the Module 16 report. Included are a description of the section, subject matter reviewed, basis for acceptance, and statement of required followup or evaluation, Section 1.0--Introduction (1) Review Introduction and Section Examinatio This section presents the scope of Module 16, lists the key hardware and work activities included, outlines the Module organization, and reports the project status. The section was examined by the NRC Region II inspectors and NRR reviewers for appropriatenes cf scope and for background informatio (2) Boundary Definitio The construc. tion area of this module addresses those activities involved with the installation of primary loop equipment only. Within the design area, this module addresses the design interface between Bechtel Power Corporation (BPC) and Westinghouse. Specific BPC design activities regarding the Westinghouse-supplied NSSS systems are addressed in other modules. Work activities considered Westinghouse generic are not addressed; however, this module addresses those Westinghouse activities considered Vogtle specific. It was difficult to define the boundaries. This was summarized in the following NRR review remarks:
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The staff has had some difficulty in understanding the criteria for including commitments from a specific area required for licensing and excluding other commitments required for licensing related to the same area (s). The applicant has stated early in the Readiness Review program that all commitments required for licensing of Vogtle would be covered in Readiness Review. However, the licensee has stated that all generic items that fall in the licensing area would be excluded from the scope of Ladiness Review due to the already redundant nature of evaluating the safety significance for generic-related commitment The problem arises from the fact that it is unclear as to what the boundaries are for generic and plant-specific items and how the licensee applied this criteria to commitment inclusion / exclusion. The potential for disagreement on what is generic or plant-specific between the staff and the licensee is substantial and could make a difference as to what type of review is performed, either by the licensee or ,
by the staff. The staff feels that the licensee should make every effort to state clearly the method used for discriminating between generic and specific items in Readiness Review Modules and the reasoning for inclusion / exclusion of commitment items that may be borderline between generic and plant specific related item No follow-up action is required by the applicant on this modul (3) Project Status. Subsection 1.3 of Module 16 gives the approximate project status for Unit 1 and common systems within the scope of this module and as of June 1,1985, as 94 percent complete for the design area and 85 percent complete for the construction area, b. Section 2.0--Organization and Division of Responsibility This section describes the organizations involved in design and construction for Module 1 It explains which personnel are responsible for specific design and construction activities, and presents the matrix organization used by the architect / engineering contractor on the projec This section was reviewed by NRC Region II inspectors for content and accuracy. In the process of this review, the responsibilities of the various organizations and personnel within the organizations were clarifie The NRC Region II inspectors had no findings in this sectio e- r
c. Section 3.0--Commitments (1) Review Introduction and Section Examinatio This section contains listings of commitments and implementing documents, which are displayed in two matrice The first matrix, entitled
" commitments," identifies all the Module 16 commitments and their source documents. The second matrix, entitled " Implementation,"
lists the document or plant feature for each commitment, as discussed in the source document, and the project document in which the commitment is implemented. The NRR review of the commitment matrix was directed toward assuring that all regulatory requirements relating to the nuclear steam supply system within the scope of Module 16 were included and properly identified. The NRC Region 11 review was directed toward verifying the proper implementation of the listed commitment (2) NRR Review of the Commitment Matrix The commitment matrix consisting of 95 commitments was measured against the Safety Evaluation Report and FSAR for compariso When applying the review methodology described above, the staff concluded that the commitment matrix (Section 3.4) agreed with most of the FSAR requirements and staff positions within the scope of this modul However, in the course of review, the staff found concerns which needed to be addressed in Section 3.4 of Module 16. There was a total of nineteen such concerns; five apparent omissions and fourteen corrections or clarification The following is a discussion of these item (a) FSAR Section 3.2.2-1, Classification of Structures, Components and Systems, Commitment No. 84 (NRC Comment) In the " Remarks" entry for this commitment in the Commitment Matrix in Section 3.4 of Module 16, it should show Table 3.2.2-1 instead of Table 2.2.2- (Resolution) Remarks entry for commitment 847 will be corrected to read: Table 3.2.2- (b) FSAR Section 5.2.1.1, Compliance with Codes and Code Cases, Commitment No. 18 (NRC Comment) " Code Cases" should be removed from the description of this commitment. Code cases are discussed in FSAR Section 5.2. (Resolution) Commitment 188 will be changed to read,
" Compliance with Code."
(c) FSAR Section 5.2.3.2.2, Compatibility with External Insulation and Environmental Atmosphere, Commitment No. 21 (
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(NRC Comment) The referenced FSAR section should be 5.2.3.2.3 instead of 5.2.3. (Resolution) Commitment 211 referenced FSAR section will be changed from 5.2.3.2.2 to 5.2.3. (d) FSAR Section 6.1.1.1, Engineered Safety Features Materials Specification Requirements, Commitment No. 32 (NRC Comment) The referenced FSAR section should be 6.1.1. instead of 6.1. (Resolution) Commitment 326 referenced FSAR section will be changed to 6.1.1. (e) FSAR Section 6.1.1.1, Engineered Safety Features Materials--Containment Penetration Materials, Commitment N .
(NRC Comment) The referenced FSAR section should be 6.1.1.1.1 instead of 6.1. (Resolution) Commitment 327 referenced FSAR section will be changed to 6.1.1. (f) FSAR Section 6.1.1.1.3.A, Engineered Safety Materials Integrity of Safety Related Components, Commitment No. 25 (NRC Comment) The referenced FSAR section should be 6.1.1.1.3 instead of 6.1.1.1.3. The entire FSAR Section 6.1.1.1.3. appears to be specific to Vogtle and all the appropriate Regulator Guides, i.e., 1.31, 1.36, 1.37, and 1.44, should be incorporate (Resolution) Regulatory Guides 1.31, 136, 137, and 1.44 have been included as commitments 1855, 1536, 1537, and 1541, respectively, in FSAR Section 1.9. However, the referenced FSAR section of commitment 256 will be changed to 6.1.1. and Regulatory Guides 1.31, and 1.36, 1.37, and 1,44 will be added for consistenc (g) FSAR Section 1.9.26 (Regulatory Guide 1.26), Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plant (NRC Comment) Regulatory Guide 1.26, is referenced in FSAR Section 3.2.2. The VEGP classification system (FSAR Section 3.2.2) is included as a commitment (No. 1754). Furthermore,
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the seismic design classification (FSAR Section 1.9.29) is included as a commitment (No. 1532). Thus, the quality group classifications should be included as a commitment for consistenc (Resolution) Nuclear Steam Supply System components are classified according to ANSI N18.2-1973 which is an acceptable alternative to Regulatory Guide 1.26. Commitment 1727 in Module 16 addresses the commitment to ANSI N1 Commitment No. 682, FSAR Section 1.9.26 which also addresses the commitment to ANSI N18.2, will be added to Module 16 for consistenc (Staff Position) Question Resolve However, GPC indicated (see response to the first question under Commitment Clarifications below) that Westinghouse will initiate a change to the FSAR Section 1.9.26 to reference both ANSI N18.2-1973 and ANSI N18.2a-1975. Thus, the added Commitment No. 682 should reference both ANSI document (h) FSAR Section 1.9.71 (Regulatory Guide 1.71), Welder Qualification for Areas of Limited Accessibilit (NRC Comment) Regulatory Guild 1,71 is referenced in FSAR Section 5.2.3.3.2. Field welds are plant specific. Thus, it appears that FSAR Section 1.9.71 is applicable to plant specific aspects of Vogtle and should be included as a commitmen Furthermore, since FSAR Section 5.2.3.3.2 is included as a commitment, the referenced Regulatory Guide should also be included as a commitment for consistenc (Resolution) As discussed in Section 1.9.71 of the FSAR, Vogtle is committed to ASME Section IX for welder qualification rather than to Regulatory Guide 1.71. This guide provides guidelines above and beyond requirements of ASME Section IX in particular to limited accessibility ASME Section IX. Few welds of limited accessibility are expected to be encountered. Reasonable engineering judgment will be used to determine if performance qualification is necessary under simulated access conditions for any identified cas Westinghouse practice does not require qualification or requalification for areas of limited accessibility as described by the guide- and this practice has provided welds of high qualit Limited accessibility qualification or requalification, which are additional to ASME Section III and IX requirements, is an unduly restrictive requirement for shop fabrication, where the welders' physical position
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relative to the welds is controlled and does not present any significant problems. In addition, shop welds of limited accessibility are repetitive due ~ to multiple production of similar components, and such welding is closely supervise For these reasons, Regulatory Guide 1.71 was not included as a commitmen (Staff Position) The staff is concerned with the position taken by the licensee on RG 1.71. The applicant has taken exception to RG 1.71 regulatory position C.1 regarding the stipulation of welder performance qualification when physical .
conditions restrict the welder's access. The alternatives proposed by the applicant have been accepted by the staff as documented in the Safety Evaluation Report (SER), Section 5.2.3. However, FSAR Section 1.9.72.2 is unclear as to what
" practice" would be followed for restricted access welding in the field. In telephone conversations with Westinghouse and '
Georgia Power, it was suggested that FSAR Section 1.9.72.2 be revised to specifically address guidelines for field welding when restricted access is encountered. Engineering judgement will be used as necessary on a case by case basis for qualification of welders working in restricted access area Georgia Power has agreed to make these FSAR changes as discussed in telephone conversations on June 23, 1986. The staff finds that this action taken by the licensee is acceptable and resolves the RG 1.71 issu (1) FSAR Section 1.9.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division (NRC Comment) Regulatory Guide 1.147 is referenced in SRP Section 5.2.1.2 and should have been included in FSAR Section 5.2.1.2. (In Commitment No.188, Code cases are incorrectly stated to be included in FSAR Section 5.2.1.1.) Acceptable Code cases evolve with time and are plant specific. Thus, FSAR Section 1.9.147 should be included as a commitmen Furthermore, since we believe that the FSAR Section 5.2. should be included as a commitment, the referenced Regulatory Guide should also be included as a commitment for consistenc (Resolution) Regulatory Guide 1.147 provides the ASME Code cases which are acceptable for use in inservice inspection (ISI) programs. ISI programs are addressed in Module 7, Plant Operations and Support, which was submitted to the NRC on January 10, 1986. Commitment 877, FSAR Section 1.9.147 of Module 7, Plant Operations and Support, addresses the commitment to the Regulatory Guid g, ,
-(j) -FSAR Section 3.2.1, Seismic Classificatio (NRC Comment) The seismic design classification (FSAR Section 1.9.29) is included as a commitment '(No. 1532).
Furthermore, the VEGP classification ' system (FSAR Section 3.2.2) is included as a commitment (No. 1754). Thus, the seismic classification should be included as a commitment for consistenc (Resolution) Module 16 addresses the commitment to Regulatory Guide 1.29 in Commitments 1532, FSAR Section 1.9.29, and 1754, FSAR Section 3. However, for consistency, Commitment 823, FSAR Section 3.2.1.1 will be adde (k) FSAR Section 5.2.1.2, Compliance with Code Case (NRC Comment) Commitment No. 188 (FSAR Section 5.2.1.1) as stated in the Commitment Matrix in Section 3.4 of Module 16 incorrectly includes Code cases. Code cases are discussed in FSAR Section 5.2. Because the applicable Code cases are specific to Vogtle, FSAR Section 5.2.1.2 should be included as a commitmen (Resolution) FSAR Section 5.2.1.2 refers to a single Code case, Code Case 1528 (SA-508, Class 2a). This Code case is being applied to other Westinghouse NSSSs with the same model steam generators and pressurizer. This Code case was considered generic for Westinghouse NSSS. Westinghouse generic design activities have been assessed by special project technical evaluations and regular QA activities, and thus are not included within the scope of Readiness Revie (1) FSAR Section 3.1.4, Nuclear Steam Supply System Components Design Classification, Commitment No. 172 (NRC Comment) FSAR Section 3.1.4 indicates that Westinghouse classifies nuclear steam supply system components according to ANSI N18.2-1973. However, FSAR Section 1.9.26.2 indicates that Westinghouse classifies components within its scope of supply using ANSI N18.21a-1975. Clarification is required for the specific referenced ANSI Code editio (Resolution) ANSI N18.2a-1975 is an addendum to ANSI N18.2-197 Both documents are applicable to Vogtle. For clarification, Westinghouse will initiate a change to the FSAR, Sections 3.1.4 and 1.9.26 to reference both document (m) FSAR Section 5.2.4, Inservice Inspection and Testing of Reactor Coolant Pressure Boundary, and FSAR Section 6.6, Inservice Inspection of Class 2 and 3 Component [~
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(NRC Comment) Preservice and inservice inspections (PSI and ISI) are plant specific and relief can be requested on a plant -specific basis. Thus, PSI and ISI should be included as commitments. However, SER Sections 5.2.4 and 6.6 indicate the need for license conditions and SER Sections 5.2.4.3 and 6.6 (PSI program) contain open items. Clarification of the assumptions regarding commitments in the area of PSI and ISI is neede (Resolutions) Preservice and inservice inspection programs are addressed in Commitments 1005, 1006, 1089, 1090, 1091, and 1092 in Module 7, Plant Operations and Support. Module 7 was submitted to NRC Region II on January 10, 198 (Staff Position) Question resolved. Module 7 includes preservice and inservice inspection program commitments that are not in the Technical Specifications (except Section 6) I and are not considered by GPC as administrative detail (n) FSAR Section 5.3.2, Pressure-Temperature Limit (NRC Comment) Because the properties of reactor vessel materials vary from plant to plant, operating limitations are based on the specific properties of the reactor vessel materials at Vogtle. Thus, FSAR Section 5.3.2 appears to be specific to Vogtle and should be included as a commitmen Furthermore, because Regulatory Guide 1.99 is referenced, FSAR Section 1.9.99 (Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials)
should also be included as a commitmen (Resolution) FSAR Section 5.3.2 Pressure-Temperature Limits, is based on the specific properties of the Vogtle reactor vessel materials. However, the Vogtle properties are used as input to analyses which are performed using the same methods and techniques used on a generic basis for Westinghouse plants. Therefore, although the inputs to the analyses and
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the results are Vogtle specific, the methods and techniques described in FSAR Section 5.3.2 are generic. Westinghouse generic design activities have been assessed by special project technical evaluations and regular QA activities, and thus are not included with the scope of Readiness Revie (o) FSAR Section 5.3.3, Reactor Vessel Integrit (NRC Comment) The applicable Code edition, the materials surveillance program and the fracture toughness of the reactor vessel materials are specific to Vogtl Thus, it appears that FSAR Section 5.3.3 should be included as a
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commitmen Furthermore, because Regulatory Guide 1.99 is r?ferenced, FSAR Section 1.9.99 (Effects of Residual Elements en Predicted Radiation Dar. age to Reactor Vessel Materials)
should al,so be included as a commitmen (Resolution) In regard to FSAR Section 5.3.3 (and the referenced Section 5.3.1), the applicable Code edition for the reactor vessel is generic to those Westinghouse NSSS.of a similar time frame as Vogtle. The fracture toughness for the reactor vessel was determined using the same methods and techniques used on a generic basis for Westinghouse NSS Therefore, these areas are considered Westinghouse generi As previously discussed Westinghouse generic design activities are not included within the scope of Readiness Review. Module 7, Plant Operations and Support, addresses reactor vessel material surveillance (FSAR Section 5.3.1.6)
in commitments 1007, 2713, and 271 (p) FSAR Section 5.4.2.5, Steam Generator Inservice Inspectio \ i (NRC Comment)
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The steam generator (SG) preservice and inservice inspections may be plant specific item '
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Accordingly, SG inspection would be included as commitments
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and the referenced Regulatory Guide would also be include ,'
s Clarificatioa of the assumptions regarding commitments in the
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area of SG inspections is neede ' '
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(Resol utio.ns) Preservice and inservice inspection programs
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Comitments 821 and 852 in Module 7 address Regulatory Guides
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1.83 and 1.121 respectivel , s .
(5;aff Position) Question resolve However, FSAR Section s'
5.'4.2.5 is not a commitment in Module 7 because it is in the
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Technical Specifications. GPC does not include commitments which are in the Technical Specifications, except for Section
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I (q) FSAR Section 6.1, Engineered Safety Features Materials, and
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(TR Comment) Some commitments in FSAR Sections 6.1 and
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, (.2(.2havebeenincludedinModule1 s However, Module 16 is fpr tne r.uclear steam supply system. Thus, it does not seem
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Further w e, cynmitments in the engineered safety features and contaiament' heat removal systems are scattered in various
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modules.' ' Thus, it is difficult for the reviewers to
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f establish consistency and determine program completeness in evaluating the systems. Clarification of the assumptions used in regard to the inclusion of these systems in Module 16 and in regard to the plant specific versus generic aspects of these systems is neede (Resolutions) Commitments from FSAR Sections 6.1 and 6. are included in Module 16 because they represent Vogtle specific . design requirements which require design interface between Bechtel and Westinghous Implementation in Module 16 verified the interface of those design requirement Commitments from FSAR Sections 6.1 and 6.2.2 are also included in other modules addressing specific design activities for the Engineered Safety Features and Containment Heat Removal System Implementation in these modules verifies implementation into the system design document (Staff Position) Question resolved. However, in future !
modules, GPC should clearly state the procedures utilized to distinguish Vogtle-specific from Westinghouse generic \
commitments, to assign commitments to specific modules and to decide which items are commitment (r) FSAR 1.9.4, Assumptions Used For Evaluating The Potential Radiation Consequence of a LOC (NRC Comment) Commitment is not consistent with SER. The applicant commits to using Regulatory Guide 1.109 dose conversion factors (DCFs) for a LOCA dose evaluation. This commitment contradicts the SER and SRP Section 15.6.5. A, which are based on conservative DCFs referenced in Regulatory Guide (Resolution) This question was reviewed by Westinghouse, and their response is:
"That the SER uses the- dose convarsion factors (DCFs)
referenced in Regulatory Guide 1.4 (ICFR No. 2, DCFs) is not explicitly stated in the SER. However, even if the SER does use the DCFs from ICRP NO. 2, this does does not constitute a contradiction between the FSAR and the SER. The analysis reported in the SER is intended to be
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an independent determination of the consequences of the LOCA, not an idsntical calculation using the exact same
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assumptions. The analysis in the SER does not come up with the same results as that in the FSAR but does yield the same conclusions, i.e., the radiological consequences of a postulated LOCA are within the exposure the guidelines set forth in 10 CFR 100.11."
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"The fact that SRP Section 15.6.5, Appendix A recommends that the staff reviewer use Regulatory Guide 1.4 when performing the staff analysis and does not make allowance for utilization of the Regulatory Guide 1.109 DCFs does not invalidate their use in the FSAR analysi The use of Regulatory Guide 1.109 DCFs is clearly indicated in FSAR Table 15.6.5-4 as a deviation from Regulatory Guide 1.4. There are many other differences between the FSAR analysis and the SER analysis. Most of these differences are a result of basing the FSAR analysis on the guidance provided by SRP Section 6. while the SER analysis is based on the guidance provided by SRP Section 15.6.5, Appendix A."
(Staff Position) Question resolved. However, the licensee has stated that use of RG 1.109 DCFs is indicated in FSAR Table 15.6.5-4 when it is in FSAR Table 15.6.5-5. The applicant must correct this in the FSA Independent analyses by the staff have determined that use of RG 1.109 DCFs for the calculation of the radiological consequences of a LOCA at Vogtle are within 10 CFR 100 guidelines. However, the DCF's in RG 1.109 are for routine releases and not intended for use with LOCA evaluations, even though the results of using RG 1.109 DCF's for LOCA analysis reached the same conclusions as the staff's independent evaluation stud (s) FSAR Section 6.2.2.2.5, Commitment No. 2406 (NRC Comment) The commitment section for Reference Number 2406 should be 6.2.2. (Resolution) FSAR section number for Commitment 2406, shown as 6.2.2.2, will be changed to read 6.2.2.2.5 in order to pinpoint the commitment location. For consistency, the other commitments in FSAR Section 6.2.2.2 will be changed to read as follows:
Reference N FSAR Section 2403 6.2.2.2.1. .2.2.2.1. .2.2.2.1. .2.2.2.1. .2.2.2.1. .2.2.2. .2.2.2. .2.2.2.2. . .
(t) Commitment No. 4484 (NRC Comment) The remarks for Reference No. 4484 should include: "Except for valves provided by NSSS supplier, which are specified for 80 percent of the nameplate rating."
(Resolution) Commitment 4484 will be changed to include the remarks column entry "Except for boric acid transfer pump and boron injection recirculation pump motors where 80 percent of rated voltage is required." To further clarify this commitment, the subject will be changed to "ESF motor specs RHR pump motor and ESF auxiliary system pump motors."
The phrase "Except for valves provided by NSSS supplier, which are specified for 80 percent of the nameplate rating"
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belongs with Commitment 4822 in Module 4 and is showr in FSAR Section 8.3.1.1. Findings The applicant has provided adequate responses to all staff questions and will incorporate the responses into the module by amendment addend The staff has some difficulty in understanding the criteria for including commitments from a specific area required for licensing and excluding other commitments required for licensing related to the same area (s). The applicant has stated early in the Readiness Review program that all commitments required for licensing of Vogtle would be covered in Readiness Review. However, the licensee has stated that all generic items that fall in the ;
licensing area would be excluded from the scope of Readiness Review due to the already redundant nature of evaluating the safety significance for generic-related commitments. The problem arises from the fact that it is unclear as to what the boundaries are for generic commitment inclusion / exclusion. The potential for disagreement on what is generic or plant-specific between the staff and the licensee is substantial and could make a difference as to what type of review performed, either by the licensee or by the staf The staff feels that the licensee should make every effort to state clearly the method used for discriminating between generic and specific items in Readiness Review Modules and the reasoning for inclusion / exclusion of commitment items that may be borderline between generic and plant specific related items. No follow-up action is required by the applicant on this module.
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Conclusion The NRR staff has reviewed Module 16, Nuclear Steam Supply System, and has concluded that all commitmentr are consistent within the defined scope of revie The Vogtle Readiness Review comments contain no omissions and correctly reflect the FSAR commitments and are thus, acceptabl (3) Region II Review of the Implementation Matrix The NRC Region II inspection consisted of a review of Subsections 3.1, 3.2 and 3.3 for content and an evaluation of 37 commitments (among the 95 listed in Subsection 3.4) for commitment implementation. Of the 37 commitments, 14 had been evaluated in Module 4. The commitments selected represented the range of commitment topics and commitment sources presented in Module 1 The examination of the sample consisted of:
o Veri fying that the Subsection 3.4 commitment matrix correlates with the Subsection 3.5 implementation matrix for a sample of the commitment o Reviewing the source document- referenced in the commitment matrix for a clear statement of requirement for each commitmen o Veri fying that the document listed in the implementation matrix properly implements the requirement as stated in the source document for each commitmen The sample of commitments reviewed and the review results are listed in Table 1 of this repor Any questions raised during the review of commitment implementation were resolve Section 4.0--Program Description (1) Review Introduction and Section Examinatio This section describes procedures and program control for design; equipment and material; material control; fabrication, installation, inspection, and testing; and turnover to Georgia Powe This section was examined by the NRC Region II inspectors for content and background information. Some of the background was useful in the review of later sections in the Module, especially Section 6.0, Program Verificatio (2) Desig Subsection provides a division of design responsibilities between Westinghouse and Bechtel in regards to the nuclear steam supply system and related area This
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subsection defines the interface between the two organizations in various disciplines e.g,. civil / structural, plant design, mechanical, nuclear, electrical, control systems, et This subsection also gives a descriptive accounting of the design proces The information proved useful for developing an understanding of the program but did not lend itself to a detailed review. The NRC Region II inspectors identified no deficiencies in the design program described in this subsection. Results from the review of Section 6 of Module 16 were used to determine how effectively the program actually functione (3) Equipment and Materials. Subsection 4.2 gives a brief description of the procurement of safety related nuclear steam supply system equipment and materials within the Westinghouse Scope of suppl No deficiencies were identified in this subsectio (4) Material Control . Subsection 4.3 described the responsibilities of the NSSS installers relative to receipt, inspection, and control of ASME Boiler and Pressure Vessel Code material within the scope of Module 16. This subsection describes the requisition process used by the responsible contractor, Nuclear Installation Services Company (NISCO), to obtain the construction material It also explains receipt inspection procedures, review of ASME Section III documentation, and materials traceability method The NRC Region II inspectors had no findings in this subsectio (5) Fabrication, Installation, Inspection, and Testing. Section describes those activities which are both pertinent and specific to the fabrication, welding, and erection of the nuclear steam supply system (NSSS) supports, and the setting, assembly, and construction testing of NSSS equipment, which are within the scope of responsibilities of Nuclear Installation Services Company (NISCO). This subsection describes program requirements such as welding /NDE, rigging and handling, and storage. Also included are brief descriptions for installation of NSSS components and related hardwar The material in this subsection is generally descriptive and was not assessed in detail by the NRC Region II
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inspector The material presented does indicate that the basic installation program addresses project requiremen Results for the review of Module 16 Section 6.0 were used to determine how effectively the program actually functioned. The NRC Region II inspectors had no findings relative to the description of this progra (6) Turnover to GP Subsection 4.5 describes the turnover of documentation for both non ASME Code work and ASME Code wor Documentation for work activities which do not require an ASME data report and certification by the Authorized Nuclear Inspector (ANI) or code stamping, is then transmitted to Georgia (GPC),
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where receipt is acknowledged by signature on the NISCO transmittal. Documentation providing objective evidence of work activities on ASME components by NISCO is retained by NISCO for compilation of the ASME data reports and subsequent reviews by the applicable signatories (e.g., ANI, Westinghouse). Upon completion of the data report and application of the NA symbol stamp, the quality documents and data report are transmitted to GPC. The NRC Region II inspectors had no findings in this subsectio e. Section 5.0--Audits and Special Investigations This section discusses quality assurance audits and evaluations that have been performed by GPC, Westinghouse, Southern Company Services, NISCO and the Institute of Nuclear Power Operations addressing pipe stress analysis and pipe support It also describes inspections and special evaluations conducted by the NRC. Section 5 discusses findings resulting from these audits relative to the design and construction programs. The section also presents six design problems considered significant enough by GPC that the NRC was informed of their potential reportability pursuant to criteria of 10 CFR 21 and 10 CFR 50.55 (e).
There were no construction problems significant enough for consideratio The findings resulting from design program audits occurred in the areas of design criteria, drawings, design documentation, and design change All of these phases of design activity were given detailed evaluation in Readiness Review's verification of the design program in Module 16, Subsection 6.1. The findings resulting from construction audits were all evaluated for importance by the Readiness Review Tea The Construction Program Verification in Module 16, Subsection 6.2, was designed to include all areas where findings of high or moderate importance occurred. The Region II inspectors reviewed the overall list of findings for these two areas, and agree that the items selected by the Readiness Review Team was representativ In addition, three internal audits performed by NISCO on Construction activities were reviewed in their entirety. There were no detectable trends indicated in the list of findings examine The NRC Region II inspectors concluded that findings from previous audits and investigations were thoroughly considered in the Readiness Review Team's assessments for Module 16. The NRC Region II inspectors had no findings in this sectio f. Section 6.0--Program Verification This section of the Module 16 Report described activities undertaken to ascertain whether design and construction work processes were adequately controlled to ensure implementation of licensing commitments and conformance with project procedures and design requirements. The
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section is divided . into two subsections . covering Design Program
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Verification and Construction Program Verification. The NRC Region II inspectors performed detailed evaluations of both subsections, which are described under the following two headings in this repor Subsection 6.1--Design Program Verification-(1) Review Introduction and Subsection Examination. . Th Design Program Verification program was performed by the Readiness Review design verification team. The design verification concentrates on the design interface 'between Westinghouse, the nuclear steam supply system (NSSS) vendor, and Bechtel Power Corporation (BPC).
The results of the review provided a basis on which to determine whether interface activities have been properly controlle Key areas were selected for review to ascertain whether the NSSS interface has been appropriately implemented and controlled. -The NRC Region II inspectors reviewed seven of the eleven areas reviewed by the Readiness Review Team. The design verification was performed in two phases by the Readiness Review Tecm. In Phase I, licensing commitments which are unique to VEGP, were reviewed to ascertain their implementation in project design document In Phase II, design documents were reviewed to ascertain whether required design interface data has been properly transmitted, received,. and implemente The NRC Region II inspectors concentrated on Phase II part of the examinatio (2) Program Examination. In Phase II, the design interface between Westinghouse and BPC/GPC was reviewed. The Readiness Review Team selected eleven key areas for review to encompass the major NSSS interface activitie The NRC Region II raview included the following seven key areas:
o Piping stress analysis o Reactor coolant loop equipment supports o Accident analyses o Instrument and control interface o Electrical interface o NSSS equipment qualification and nozzle loads o Equipment installation requirement The NRC Region II inspectors examined samples of design interface items in each of the key areas. These samples included data provided to Westinghouse by BPC and data provided to BPC by Westinghouse. The review was directed as much as possible toward
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changes in design which would result in interface activities. To-accomplish this, the review also included samples of systems and components for which design changes have occurred that required additional design interface activitie The Westinghouse /BPC/GPC interface activities were examined for proper and effective exchange of information between
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organizations. The areas considered were transmission of information required by other parties, receipt and correct-interval distribution of the information, implementation of information, and feedback, when neede The following sections discuss the review and results of the selected interface area (a) Pipe Stress Analyses The Bechtel/ Westinghouse interface in t.he pipe stress analysis area was reviewed for proper control and. effective exchange of data required by each organization and feedback of information. The review of the interface in this area concentrated on design changes which required interface activity. The following areas were selected for review to encompass most of the interface activities in this area:
o Jet impingement loads on pressurizer surge line o RHR Recirculation and Hot Leg Injection o Support load changes o Support location change Design changes were selected for review and additional samples were included to ensure a review in each of the above area ) Jet Impingement Loads -The following documents were reviewed:
o B-W 4358, dated August 3, 1984, this Bechtel document transmitted the complete set of jet impingement load o B-W 4080, dated August 9, 1984, this Bechtel document transmits pipe insulation dat o W-8, GP 7483, dated August 30, 1983, surge line data accepted, except for some support stiffnesses.
- o W-B, GP 8690, dated October 17, 1984, transmits results of set impingement evaluations
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o W-8,GP 8621, dated September 13, 1984, suggests revised snubber size for unit 2 o- DWG 1X4DL4A17, R/4, Containment Bldg Piping Areas 4A, B, C, D, Level B - Class and Sections Reactor Coolant Loops The pressurizer surge line was appreciably addressed in each transmitta ) RHR Recirculation and SIS Hot Leg Injection and Stress Analysis Input -The interface in the pipe stress analysis area was observed to be properly controlled and data effectively exchange o B-W 4031, dated June 16, 1983, transmitting piping analysis Loop 4, RHR & SIS Hot Leg o B-W 4042, dated June 28, 1983, transmitting piping analysis Loop 1, RHR & SIS Hot Leg Also 3 transmitted were 3 isometrics for the above items:
1K4-1204-016-01 R/2 1K4-1204-049-02 R/1 1K4-1204-197-02 R/2 3) Support Load Changes o W-B, GP. 9228, dated March 26, 1985, transmits analysis results SIS Hot Leg Loops
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Attachment I-Support Design Load 5 and Displacements
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Attachment II- As analyzed support configuration o Supports
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VI-1204-025-H001 R/1 - Rigid Strut and DCN-1
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VI-1204-025-H002 R/1 - Rigid Strut
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VI-1204-025-H002 R/4 - Snubber
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VI-1204-025-H012 R/1 - Snubber
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VI-1204-047-H003 R/3 - Snubber and DCN-1
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VI-1204-047-H003 R/3 - Snubber 4) Support location change o W-8, GP 8916, dated December 28, 1984, advises BPC of support load and configuration change o ISO 1K4-1204-051-07 o Hanger V11204-051-H028 R/1 and R/2, OCNR-1 (b) Reactor Coolant Loop Equipment Supports. The review of the interface in the area of primary equipment support loads on the ~ containment structure concentrated on the changes._in the loads for the steam generator supports. Westinghouse again revised the support loads for this equipment in April 198 These new loads were transmitted to BP The following documents were reviewed:
o W-B, GP-9325, dated April 23, 1985 transmitted Support Analysis WCAP-10734 Vo o BPC calculation X2CJ4.2.2 R/3 Steam Generator Support Embeds o BPC calculation X2CJ4.2.2 R/4 revising Steam Generator Support Loads The interface in the area of primary support loads on the
. containment structure and the revision of loads were observed to be documented and controlle (c) Accident Analysis Westinghouse performs several types of accident analyses which require information from Bechte The containment pressure / temperature calculation data was selected for revie The references from Westinghouse requesting data necessary for performing the pressure / temperature calculation are:
o Westingnouse letter GP-3602, dated January 25, 1980 o Westinghouse letter GP-3623, dated January 31, 1980 A review was made of the data requested in the letters compared to the data supplied in letter designated as Bechtel log No. BW2999 dated August 15, 198 The required information was supplie .. .
Additional information was supplied to Westinghouse by Bechtel after the VRR team review for Module 16. The following two letters transmitted the additional or changed data:
o Letter Bechtel Log No. 4647 dated July 8,1985 o Letter Bechtel Log No. 4655 dated July 31, 1985 g (d) Instrumentation And Control The interface in the instrument and control area was reviewed for incorporation of functional requirements specified by Westinghouse for Bechtel-designed systems, and for the input signals to be provided by Bechtel to the Westinghouse solid state protection system (SSPS), which is a part of the engineered safeguards features actuation syste The following were reviewed:
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Function Requirements The Westinghouse functional requirements which were to be implemented by Bechtel in the instrumentation and control area were reviewed using as an example the blackout signal for start of the turbine driven auxiliary feedwater (AFW) pump. The following drawings were reviewed:
o Drawing No. 1X4DB161-3 Rev. 14 P&I Diagram, Auxiliary Feedwater Pump System (Auxiliary Feedwater Pump Turbine Driven) System 1302 o Drawing No. 1X5DN120-2 Rev. 4 Control Logic Diagram Auxiliary Feedwater System A review of these two drawings indicated that a blackout signal would initiate the start of the turbine driven AFW pump and satisfies the requirement in Westinghouse document GAE/GBE-300/7 Rev. 3
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Solid State Protection Systen Input Signals an instrumentation and control input signal for the main steam stop valve position required to be supplied by Bechtel to the Westinghouse Solid State Protection System were verified by reviewing the following drawings:
o Drawing No. 7243007 sheet 16, Logic Diagram (Westinghouse)
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o Drawing No. 1X39-CD-B14A Rev.6, wiring Diagram Solid State Protection Cabling Block Diagram (Bechtel).
(e) Electrical ,
Samples of the interface in the electrical area was reviewed in three areas: engineered safeguards features, actuation system (ESFAS), valve and pump train assignment Westinghouse electrical requirements for Bechtel-designed systems, and input signals to be provided by Bechtel to the Westinghouse SSPS which is a part of the ESFA Train Assignments Two pairs of valves and the two safety injection ptmps were reviewed and verified as having the proper train assignment by reviewing the following drawings Valve or Pump N Drawing N Rev. N Valve 8801A 1X3D-BD-D02E 6 Valve 88018 1X30-BD-D02F 5 Valve 8809A 1X3D-BD-D02V 5 Valve 8809B IX3D-BD-002W 5 SI Pump 1 1X39-BD-D01C 3 SI Pump 2 1X39-BD-D01D 5
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Power Requirements The AFW turbine driven pump flow path was verified as being independent from the diesel AC power source (a Westinghouse Steam Systems Design Manual requirement).
A review of the following three drawings verified that motor operated valve 1 HV-5106, which when opened allows steam to drive the turbine driven AFW pump, was dependent on 125 volt DC power:
o Drawing No. 15DN120-2 Rev. 4, control Logic Diagram Auxiliary Feedwater Syste o Drawing No. 1X4DB161-3 Rev. 14, P & I Diagram Auxiliary Feedwater Pump System (Auxiliary Feedwater Pump Turbine Driven)
o Drawing No.1X3D-BC-F02A Rev. 3, Elementary Diagram Auxiliary Feedwater Syste s. . ,-
Also, the independence of lube oil pump power and the lube oil cooling source from the diesel AC power source were verified by reviewing the following two drawings:
o Drawing No. C-6HMTA 321X88 (Bechtel Log N X4AF03-130(1)-1)
o Drawing No. 103323E Sheet 1 of 2 from the Vendor Manual (Terry Corp.) Bechtel Log No 2X4AF03-229-2
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Solid State Protection System Input Signals and electrical input signal required to be supplied by Bechtel to the Westinghouse Solid State Protection System was verified by reviewing the following four Bechtel drawing:
o Drawing No IX3D-BD-B01N o Drawing No. 1X3D-BD-801P o Drawing No. 1X3D-80-801X o Drawing No. 1X3D-BD-B01Y The reactor coolant pump under frequency signal was shown as a separate channel on each of the drawings showing the signal had been implemented appropriatel (f) NSSS Equipment Qualification and Nozzle Loads Nuclear steam supply system equipment qualification was reviewed for nozzle loading on equipment from the attached piping. The sample selected for review was the Steam Generator No. 1, Main Steam Nozzle loading. For the steam generator, the allowable nozzle loading specified by Westinghouse E-S Specification N , Rev. 1, transmitted to Bechtel by GP-5121 dated 8/13/8 This document was referenced in, Bechtel Log No. IX6AB12-32-1.,
received 8/25/81. The NRC Region II inspectors also reviewed Bechtel calculation X4CP-7110 R/2, page 10A/25A, which shows the steam generator nozzle loads and allowable The reference given in the calculation was; "Given by W Table 2 of Specification X6AK03-46-3". This is an incorrect reference. Also in the Fc Direction for the weight condition the allowable in the calculation was stated as (+ or -) 15 kips. This is also incorrect, the proper allowable is (+ or
-) 10 Kip In as much as the Readiness Review Team's Checklist, Item H-8 for the Steam Generator Main Steam Nozzle loads was checked in this calculation and the Readiness Review Team did not identify the discrepancie This deficiency is identified as Finding No.1 and 2 in section 4 of this repor '
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(g) Equipment Installation Requirements The interface in the area. of equipment installation requirements was reviewed in the control systems are Process Control Instrumentation was selected for review. The sample .for review was selected to include various types of Class IE transmitters and to include installation. The NRC Region II inspectors selected the RWST Level and pressurizer pressure transmitters for revie It was during.this portion of the review that the Readiness Review Team identified that the torque requirements for the transmitter bracket mounting bolts, shown on.the Westinghouse drawing, were not included in the Bechtel drawing. Another example is the transmitter bracket mounting bolts shown on the Bechtel drawing as 1/4" bolts, not the 5/16" bolts shewn on the Westinghouse drawing. This was identified by the Readiness Review Team as Finding 16-13, which is discussed later in this repor The following documents were reviewed:
Bechtel Drawing N Westinghouse Transmittal Letter 8765D52 R/3 GP-7029 8765D66 R/6 GP-7611 8765067 R/5 GP-9316 8765D68 R/3 GP-2601 8765D69 R/6 GP-8327 The installation requirements from Nuclear Instrument System (NIS) cable were selected for review. Westinghouse provides the installation requirements in the NIS Cable and Connectors Installation Control and Electrial System (C & ES) standard documen This information was referred to in Bechtel document, Construction Specification X6AS01- The following documents were reviewed:
o C & ES Standard R/5, transmitted by GP 3747, dated May 20, 1980 o Drawing 1X3DF332 R/5 o Bechtel Construction Specification X3AR01, Section E8, R/21 In the areas reviewed, except as noted above, the Westinghouse installation requirements were reflected in Bechtel design document ~-
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(h) ' Findings -
Subsection 6.1.4 of ' the Module 16 Report presents four findings disclosed by the Readiness Review Team in the Design
! Program Verification. Among these findings were-one Level I, two Level II, and one Level III findings. The NRC Region II inspectors performed a detr.iled evaluation on the one Level I finding and is described below:
o Finding 16-13--This firding noted that the installation
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bolting details, bolt size and -torque valves per
Westinghouse supplied seismically and environmentally qualified electronic DP transmitters and pressure transmitters are not-the same on the Bechtel drawings as those on the Westinghouse drawings. The Readiness Review Team concluded that in the initial issue of the
- Instrument Installation drawings, certain vendor requirements for instrument installation were inadvertertly omitted from review and were- not incorporated. The inspectors concluded that all changes would be properly incorporated in the documentation under the program identified for corrective actio The examination of the Readiness Review Findings (Level I) in the Design Program verification disclosed no verification error Subsection 6.2--Construction Program Verification (1) Review Introduction and Subsection Examination. The Construction Program Verification performed by the Readiness Review Team consisted of an evaluation of the installation of the nuclear steam supply system (NSSS) and related construction activities performed by Nuclear Installation Services Company (NISCO). The program had the objective of determining whether the construction control process functioned effectively and whether it insured acceptable installation of the NSSS component (2) Program Examination. The assessments made and conclusions reached in Readiness Reviews verification program were divided into the two following major categories
o Hardware / Components o Programs / Procedures A further breakdown of the areas and attributes appraised is as follows:
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i (a) Hardware / Components
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- The - installation ~ activities performed by NISCO for the
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-following components were chosen:
Reactor coolant pump'No. 4 motor settin Reactor coolant pump No. 4 suppport columns and tie
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- Steam generator No. Pressurize Reactor pressure vessel (RPV).
- RPV head assembl .RPV internal Bottom-mounted instrumentation:
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A ' list of' attributes and activities examined for the installation of these components were:
- Documentation-of activities
- RiggingLand lifting
- Location and ' orientation
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- Material / Component identificaticn
- Clearances-
- Welding control and inspection This is a composite list and not_every component was examined for each of the attribute For example welding was not performed by NISCO on all of the listed component (b) Programs /Procedurg Programmatic activities examined that supported _NISCO's field installation activities were:
' - Material Control,
- Document / records control,
- Personnel certifications,
- Nonconformance handling, j A list of attributes examined under each of the programmatic
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activities were:
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Materials Control i-
.. - Identification of the material (type, size, tag no.,
heat no);
- - Appropriate signatures for inspections;
- CMTRs. if required, in conformance with specifications;
! - Evidence of acceptance or rejection.
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. Nondestructive Examination-Test reports were reviewed for:
- Test part identification;
. Consumable materials;
- Governing code;
- Technique data;
- Examiner's qualification;
- Examination results;
- Legibility;
- Completenes Documents / Records Control
- Drawing numbers listed on the PCS were correct for the time of installation / inspection;
- Drawing revisions were documented on the PCSs;
- PCSs reflect appropriate information and entries;
- PCSs were available and retrievabl Personnel Certifications
- Review of QC inspectors and NDE technicians qualifications to ascertain that QC personnel were qualified to the appropriate level for the-QC test method being use Nonconformance Handling
- Appropriate disposition approval signatures;
- Proper closure and completio (3) Program Examination by Region II The Region II inspectors selected three NSSS hardware / components installed by NISCO. These were installation of steam generator n I, pressurizer, and bottom mounted instrumentatior guide tube While the Readiness Review Team divided the program examination into two major categories, the Region II inspectors combined these two categories for the purpose of improving technical effectivenes (a) Installation of Steam Generator No. 1 NISCO writes engineering specifications for installation of NSSS components. These specifications are then submitted to Bechtel and Westinghouse for approval. After the approval is received by NISCO then a working or installation document
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called a Process Control Sheet (PCS)'is prepare The PCS has step by step instructions, QC sign off, ANI holdpoints, drawing numbers and revisions, receiving inspection report numbers, et NISCO engineering specification ES-4028-3, Steam Generator
~ Supports and Final , Setting Procedure, was compared with the following process control sheets (PCS) to ascertain that all of the requirements stated in the specification were entered on the PCS's:
PCS No. 120-1 for Setting the Steam Generator PCS No. 958-13 for Steam Generator Vertical Column Supports PCS No. 958-13-1 for Steam Generator Vertical Column Supports l PCS No. 958-25 for Steam Generator Upper Lateral Supports PCS No. 958-25-4 for Steam Generator Upper Lateral Supports PCS No. 958-25-6 for Steam Generator Upper Lateral Supports PCS No. 958-25-7 for Steam Generator Upper Lateral Supports It was confirmed that the requirements were contained in the PCS's, that approved procedures were used, that approved hold points were established and observed, and that installation had been properly recorded and subsequently reviewed by NISC The steam generator supports were supplied by Westinghouse and it was not necessary for NISCO to perform any welding for the erection of the supports. A visual examination was performed on the quality and sizes of the welds on the supports versus the drawing requirements. The components of the supports were identified as depicted on the drawing The supports for steam generator No. I were received on
receiving inspection report (RIR) No. 5 and the documentation for the materials in this RIR were reviewed and accepted by NISCO. During the visual inspection of the supports, several i components were noted for material acceptance verificatio The following materials were selected:
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Component and Westinghouse I
Serial N Heat N Material Vertical Column S/N201-1 DWD A-588 DYF A-588
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- S/N202-2 DWB A-572 i
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S/N6635 '12-6935 A-471 S/N6623 22-5017 A-471-The certified materials test reports met the applicable material specifications. Also,. materials A-588, A-572, and A-471 were code case No. 1644 ~ materials which met the requirements of Regulatory Guide 1.85 Rev, 24 dated June 198 The following deviation reports (DR) were reviewed:
DR N Subject NI-00003 Part not on approved vendor lis NI-00066 Vertical column out of drawing toleranc MO-3559 Damage to steam generator support The deviations were ascertained to have the proper disposition, completion, and closur (b) Installation of Pressurizer Installation of the pressurizer involved a ring plate, 24 bolts with nuts, and washers. In addition, NISCO installed the pressurizer seismic restraints that had been fabricated by Teledyne Brow NISCO engineering specification ES 4028-5, Pressurizer Final Setting Procedure, also contains information for erecting the seismic restraints and was compared with the following PCS's to ascertain that all of the requirements stated in the specification were entered on the PCS's:
PCS No.130-1 for Setting the Pressurizer-PCS No. 958-33 for Installing Seismic Restraints PCS No. 958-33-1 for Shims for Seismic Restraints PCS No. 958-33-2 for Locking Bars for Seismic Restraints PCS No. 958-33-3 for Arc Strikes on Seismic Restraints It was confirmed that the requirements were contained in the PCS's, that approved procedures were used, that approved hold points were established and observed, and that installation had been properly recorded and subsequently reviewed by NISC No welding was performed by NISCO for the erection of the steam generator and seismic restraint However, a visual inspection was performed on a sample of the welds for the seismic restraints. The weld met the quality required by the specifications. The sizes specified on the drawing had to be altered in several places to allow the installation of the L
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31-bolting. This condition and the dispositions were documented on_a deviation report. An arc strike had been removed from lateral support (restraint) No. 801-1 and the area was dye penetrant inspected again for this evaluatio The following DR's were reviewed:
DR N Subject NI-00008 Arc strikes on lateral supports NI 00030 Modification of welds, washers, and nuts to facilitate installation of lateral supports The deviations ha6 the proper dispositions, completion, and closur (c) Installation of Bottom Mounted Instrumentation Guide Tubes and Support The bottom mounted instrumentation guide tubes run from the reactor vessel to a seal table and are part of the reactor coolant pressure boundary. The installation of the guide tubes and their supports involved ASME Code welding by NISCO while the installation of the steam generator and pressurizer did not. The relevant ASME Code for the - tubes is ASME Section III Subsection NB Class 1, and for the supports is ASME .Section III Subsection NF Class 1. The program examination on this hardware was first performed on the guide tubes and then on their support Installation of Guide Tubes NISCO engineering specification ES-4028-12, Bottom Mounted Instrumentation Guide Tube Assembly, was compared with the following process control sheets (PCS) to ascertain that all of the requirements stated in the specification were entered on the PCS' PCS No. 90-1 for Guide Tube Installation PCS No. 90-1-6 for Reworking Guide Tube Thimbles PCS No. 90-1-31 for NDE of Seal Table Welds It was confirmed that the requirements were contained in the PCS's, that approved procedures were used, that approved hold points were established and observed, and that installation had been properly recorded and subsequently reviewed by-NISC The following documentation for weld material was compared with the ASME Code requirements:
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Weld N Filler Metal Type Heat N RIR 1A9 ER Ni Cr-3' A5006N382 113 1A9 ER Ni Cr-3 C4972G382 113 2F1 ER 308 L 26245 61 2F1 ER 308 L 05394 61 The filler metal met the ASME Code requirement The following documentation for other material in the guide tubes was compared with the ASME Code requirements:
Component Material Alloy- Heat N RIR N Seal Table SA240 304 817925 172 Guide Coupling SA403 304 56873 57 Guide Tube No. 55N14 $A213 304L 465013 57 Guide Tube No. IJ8 SA213 304L 463582 57 The material met the ASME Code requirements There were 58 guide tubes and NISCO performed six welds on each tube from the reactor vessel penetration to and including the seal table weld. The following NISCO welding procedures used for performing these welds were reviewed:
WPS 438.2.2, Welding Procedure Specification for Manual Gas Tungsten Arc Welding (GTAW) of Inconel Bottom Head Adapters to Stainless Steel Instrumentation Tubing (Reactor Bottom Mounted Instrumentation Connections).
GWP-BMI-1, General Welding Procedure-Reactor Bottom Mounted Instrumentation Connection WPS 80.2.5, Welding Procedure Specifications for Manual Gas Tungsten Arc Welding (GTAW) of Austenitic Stainless Steel Couplings and Seal Plate to Austenitic Stainless Steel Instrument Tubing (Reactor Bottom Space Mounted Instrumentation Connections).
Engineering Specification ES56, Welding Filler Metal Control Procedur The PCS's were examined for determining if visual, fit up and dye penetrant inspections had been performed for the following welds:
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Tube N Weld No Welder B-6 1 thru 6 N-15 C-7 1 thru 6 N-12 D-10 1 thru N-20 E-5 1 thru 6 N-15 F-1 1 thru 6 N-19 The PCS's indicat.ed that the quality control inspections had been performed. Welder qualifications and ASME Code welding continuity _ records verified that the Welders were qualified to make these weld The - PCS's were also examined for x-ray (RT) inspections performed on weld No. 1 (reactor vessel penetration to instrument guide tubes). The following sample was taken:
Accepted Tube N RT Report N B-6 67 D-10 125 F-1 126 In the area of NDE the following NISCO procedures were reviewed:
- Engineering Specification E. S.100-5, Visual Inspection of Welds Engineering Specification , Liquid Penetrant Examination During a walkdown of the bottom mounted instrumentation guide tubes, a random sample of the welds were visually inspected and dye penetrant inspected again. Weld No. 1 for 30 of the 48 guide tubes was visually inspected by the Region II inspectors with acceptable result However, a mechanical impression was found on tube No.48-P-4 approximately one foot below weld No. 1. Evaluation of this potential tube damage was not documented by NISCO. This was identified to the licensee as one of two examples of a deviation (reference VIOL 424/86-78-02).
Visual examination was also performed on a sample of seal table welds during the observance of dye penetrant inspection by a NISCO QC inspecto Dye penetrant inspections requested and observed by the Region II inspectors are as follows:
. -- ._
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Tube N Weld N J-1 -1 30-F-3- 1 28-C-7 1 35-D-8 4
~4-H-6 5 57-H-6- 5 Weld Nos. J10, B6, and H3 on the seal table were also dye penetrant : inspected. The tests were performed and evaluated to the NISCO engineering specification and ASME Code requirement '
The NDE certifications for the inspector who performed the test, and for two other inspectors found on the PCS's were examined and found to be qualified to the appropriate levels for the QC test method being use The.following OR's were reviewed:
DR N Subject NI-00006 Lead-Copper go no go gage used on tubing NI-00014 Tube end prep damaged NI-00016 Linear indications on coupling NI-00085 Seal plate hole too small NI-00104 Wrong tube to seal plate hole The - deviations had the proper disposition, completion, and closur Installation of Guide Tube Supports NISCO engineering specification ES-4028-13, Installation of Bottom Mounted Instrumentation Guide Tube Supports, was compared with the following PCS's to ascertain that the requirements stated in the specification were addressed on the FCS's:
PCS No. 90-2 for Support 1 (Group A) installation PCS No. 90-3 for Support 2 (Group B) installation
,
PCS No. 90-4 for Support 3 (Group C) installation PCS.No. 90-6 for Support 5 (Group E1) installation It was confirmed that the requirements were contained in the PCS'S, that approved procedures were used, that approved hold. O points were established and observed, and that installation had been properly recorded and subsequent reviewed by NISC r
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The following documentation for weld material and welders who performed welding on the supports was examined to determine if ASME Code requirements were met:
Weld N Filler Metal Heat or Lot N Welder BM181 E-7018 422X8361 N12 BM1B144 E-70S2 97401 N26 BM18507 E-70S2 97401 N26 and N36 BM18876 E-70S2 97401 N26 BM1B1154 E-70S2 97401 N26 The weld filler metal met ASME Code requirements, and the welder qualifications and continuity records verified that the welders were qualified to make these weld The following documentation for other material used in the supports was compared with ASME Code requirements:
Material Heat N RIR N SA-36 422P8311 82 A-588 411N0341 102 SA-307 KC3682 140 SA-36 422T5131 154 The material met the ASME Code requirement The following NISCO weld procedure used for making the support welds were reviewed:
WPS 10.1.6, Welding Procedure Specification for Manual Shielded Metal Arc Welding (SMAW) of Carbon Steel to Itself on plate, pipe, tube, etc., 3/16" thru 5/8" thicknes WPS 10.3.1, Welding Procedure Specification for Gas Tugsten Arc Welding (GTAW) for Carbon Steel to Itself on Plate, Pipe, Tube etc., 1/16" to 5/8" thicknes A walkdown of some of the supports for the BMI tube guides was performed in order to visually inspect some of the welds and some of the dimensional attributes for the support Support 1 (Group A) the first support starting from the reactor vessel and going towards the seal table had 1153 welds. A visual and dimensional inspection (fillet size) was performed on a random sample of 43 welds. Visual inspections were performed on three other supports in this system with the only apparent deviation noted on the support E-2. Two
"U" bolts attached to the support structure steel for the purpose of restraining the guide tubes appeared to have larger clearances between the "U" bolt and guide tubes. Tube
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A No. 49-D-14; had > greater than J 1/8"- clearance and tube N D-3 had greater;than 1/4'f clearance. Westinghouse drawin No.1599E75 -sheet 2.-(Bechtel' Log No. 1X6AB08-34_-8) showe '1/16" each side or total of _1/8" allowed clearance. :This-
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deviation -from the drawing had not been documented by NISC :-This was identified .to the . licensee as ~ the -second of two-examples of a deviation'(reference VIOL 424/86-78-02).
.The followin'g D. R.~'s were reviewed:
DR N Subject NI-00070 Liner indications where. angle clips remove NI-00057 Missing hardware from BMI support steel NI-00107 -Carbon content exceed chemistry requirement NI-00009 "U" bolts for supports undersiz The deviations had the proper disposition, completion -and closur (e) Findings Subsection 6.2 of the Module 16 Report presents five_' findings disclosed by the Readiness Review Team in the Construction Program Verification. Four of the findings were in the paperwork category and one finding was in the programmatic-category. There were no findings in the hardware categor ~
-In the level of importance for the findings, there were no level I, there were three level .II and two level III-finding The NRC Region II inspectors performed an evaluation on all five findings, but performed a more detailed review of the following two findings:
o Finding 16-6 This finding noted that the assembly sequence for the reactor pressure vessel internals was
. not in accordance with specification, and that the.
,
flange leveling requirements had not been implemented.
! A review of the response indicated that the- assembly activities would not affect final internal assembly and
- installatio The upper internals assembly flange
- levelness had not been checked at the time of the Module 16 inspection. A later check found the levelness to be within tolerance. A Deviation Report for the two items
'
was dispositioned Use-As-Is.
o Finding - This finding noted four minor documentation discrepancies with the radiographic report None of
- these discrepancies questioned the quality of the weld.
l A fifth item of the finding involved five films of a
- similar joint configuration that displayed linear indication not noted on the radiographic report The
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linear indications were not addressed because they have been interpreted to be slag pushed up under the clip angle.and out of the weld area, which is common with this joint configuratio Other ASM Code film was re-reviewed to ensure that-linear indications had been addressed and the x-ray film interpretation was found to be -in full compliance with Code requiremen The examination of the Readiness Review Findings in the Construction Program Verification disclosed no verification error . Section 7.0--Independent Design Review (1) Review Introductio This section describes the Independent Design Review (IDR) of design criteria, calculations, specifications, and drawings that was performed by Stone and Webster Engineering Corporation. The intent of the IDR was to ensure that proper design requirements were considered and that implementation of design commitments was technically acceptabl '
(2) IDR Examinatio The IE examination of the IDR will focus on determining whether a thorough review had been performed and whether conclusions were appropriate relative to the findings reported. This will be accomplished during the review of Module 22, " Independent Design Review." Section 8.0--Program Assessment / Conclusions This section presents a summary of open corrective actions on Readiness
,
Review findings, several certifications and acceptances of the Module, and brief resumes for the Readiness Review Team members. The section
'
does not contain any introductory explanations as to the significance of the information presented, which is somewhat centrary to the previous sections of the Module.
l The NRC Region II inspectors' review of this section included reading ,
I for content and background information. The open corrective actions identified therein were evaluated for completeness and consistency with the finding descriptions given in Module 16, Section No findings were identified in the corrective action . Findings The following three findings were identified during the NRC evaluation of Module 16. All of the deficiencies noted are considered to have potential safety significance at this point of review and should be evaluated further i
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to preclude safety problems. Two of . the items have been identified as violations and one has been identified as a deficiency item. These will be addressed by the NRC during the routine inspection program, Deficiency - (VIOL 424/86-78-01) - Incorrect steam generator main steam nozzle load allowables and reference specification. Two deficiencies were noted in the design calculations for the steam generator main steam nozzle allowables. Details for this item are provided in 3.g(6)
of this repor Deficiency - (DI 424/86-78-03) - Inadequate review of calculation for steam generator main steam nozzle loads. The Readiness Review Team reviewed the same area and documents indicated in the previous deficiency and did not detect the error. Details for this item are provided in 3.g(6) of this repor Deficiency - (VIOL 424/86-78-02) - Failure to follow procedure for documenting deviations. Two examples - failure to document damaged area on a bottom mounted instrumentation guide tube and a clearance which did not conform to drawing for a support. Details for this item are provided in 3.h (3)(c) of this repor . Conclusions Based upon the review within the scope of this module, the NRC has reached the following conclusions for the nuclear steam supply system for Vogtle Unit Summary of Specific Conclusions With the exceptions of those items / areas discussed earlier the following has been determined to be acceptabl (1) Boundaries - The content of Module 16 and the boundaries between other modules as given in Section 1.0 were reviewed. It was pointed out in the NRR review that it was unclear as to what boundaries are for generic and plant-specific items and how the
- licensee applied these criteria. It was suggested that in future modules more information for making this distinction be included.
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(2) Organization - The description of the organization and
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responsibilities for project activities as given in Section 2 was reviewed and found to be accurate and to include the pertinent activities for design and field construction.
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(3) Commitments - The commitments as given in Section 3.0 were reviewed and determined to be complete with exceptions for i commitr 2 Nos.1727 and 2021 which will require changes to the FSAR, the licensing commitments and implementing documents were determined to be in compliance with the FSAR, the SRP, Regulatory Guides, and industry codes and standards.
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Ng (4) Program - The description of the program as given in Section ' "
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was reviewed. The following areas were examined:
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Design wy y s
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Equipme.,t'and Material Material Control Y,'. l > Q1.'
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't & Fabrication, Installation, Inspection, and Testing
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A Turnover to GPC
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The program description was determined to be generally correct and
'] in agreement with FSAR and requirement b (5) Audits and Special Investigations - Section 5.0 on audits was
% reviewed and' determined to be an accurate presentation of the m , audit process and,previously identified construction problems and NRC inspection result s-1 r s
.
i (6) Program Verificatloa - GPC performed a program verification in two parts; a design program verification and a construction program
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verification. The design program verification concentrated on the N desiga interface (i.e. the flow of design information) between Bechtel Power Corporation and Westinghouse. Eleven -key nuclear
-
s[ 3 4teamJsu,pply system (NSSS) areas were involved in the verification c sample. 'The design review was conducted in two phases. In Phase
,e % c I the licensing commitments which are unique to VEGP (i.e. not
- Westinghouse NSSS generic design) were reviewed to ascertain their implementation in project design documents. In Phase II design documents were reviewed to ascertain whether required design interface data has been properly transmitted, received, and implemented. The NRC Region II inspectors sampled 7 out of the 11 NSSS areas from Phase II and found the review that was performed by the Readiness Review Team was acceptable.
,
The construction program verification section of this module was performed to provide an evaluation of nuclear steam supply system installation and related construction activities performed by Nuclear Installation Services Company (NISCO). The assessment l areas selected were divided into two parts, hardware / components a.1d program / procedures. The NRC Region II inspector reviewed the
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' construction program verification activities for the installation
.cf a steam generator, the pressurizer, and the bottom mounted instrumentation guide tube The review performed by the
/\' ,f Readiness Review Team was acceptabl s
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'(7) Independe.it Design Review - GPC contracted as outside (s , o r'gani zation , Stone and Webster Engineering Corporation, to ti perform an independent Design Review (IOR) for Module 16. This
,s review was performed to assess the technical adequacy of the
' Module 16 related design for Vogtle The IE examination of the
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performed and whether conclusions were appropriate relative to the findings reported. This will be accomplished during the review of Module 22, " Independent Design Review."
'
(8) Program,- Assessment / Conclusions - This section presents a summary of open corrective actions on Readiness Review findings, several W certifications and acceptances of the Module, and brief resumes for the Readiness Review Team members. The section does not contain any introdestory explanations as to the significance of the information presented, which is somewhat contrary to the ('
~
previous sections of.the Modul =
The NRC Region II inspectors' review of this section included reading for content and background informatio The open corrective actions identified wherein were evaluated for completeness and consistency with the finding descriptions given in Module 16, Section No findings were identified in the corrective action The examination performed by NRC indicated that this module presents
~ and adequate assessment of the GPC process for design, fabrication, installation and inspection of the nuclear steam supply system for Vogtle GPC's management supported the Readiness Review by active participation and adequate resources. There was no evidence of coercion or attempt to dilute either the effort or the finding The Readiness Review Staff displayed the requisite, competence and professionalism for a review of this natur The review performed by the Readiness Review Staff was determined to be sufficiently comprehensive in scope and depth to identify problem areas, and the dispositions of findings determined to be adequate. The procedures for -design, engineering, construction, and quality control for the applicants program was comprehensive and provides adequate assurance that activities associated with the nuclear steam supply system were determined to be consistent with commitments and are acceptable. Based on the review of this module, it appears that construction was performed in accordance with the appropriate procedures and that records reflect the quality of that constructio With exception of the three deficiencies reported in Section 4, the NRC findings appear to have minor significance and do not appear to represent significant programmatic weaknesses. The deficiencies must be further evaluated to determine their significance and need for any subsequent corrective actions.
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_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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Pending resolution of the findings identified above, the NRC ' concludes Lthat the Vogtle -program for the nuclear steam supply system complies with NRC requirements and FSAR commitments. This conclusion is based on information currently available to the inspectors and reviewer Should subsequent contradictory information become available it will be evaluated to determine what effect 'it may have. on the above-conclusions.
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COMMITMENT VERIFICATION ,
Re fe rence Commitment Commitment Responsible Number Sou rce Section Commitment Sub iect Document Fea t u_ re Section 1532 FSAR 1.9.29 Seismic Design Classification RG 1.29, Rev. 3, Design 9/78 1855 FSAR 1.9.31 Control of Ferrite Content in RG 1.31, Rev. 3, Design Stainless Steel Wold Metal 4/78 1536 FSAR 1.9.36 Non-Meta l l ic The rma l Insulation RG 1.36, Rev. O, Design for Austenitic /73 1537 FSAR 1.9.37 QA Requi rements for Cleaning of RG 1.37, Rev. O, Design fluid Sys. & Assoc. Components (Ha rch 73)
1538 FSAR I.9.37 QA RequirementG for Cleaning of ANSI N45.2.1-1973 Design and Fluid Systems & Assoc. Components Construction 3201 FSAR 1.9.37 Quality Assurance Requi rements for RG 1.37, 3/73 Design and for Cleaning of Fluid Systems and Construction Associated Components of Water-Cooled Nuclea r Power Plant (1.9.37.2)
1541 FSAR 1.9.44 Control of Use of Sensitized RG 1.44, Rev. O, Design 5/73 1546 FSAR 1.9.50 Control of Preheat Temps. for Welding RG 1.50, Rev. O, Design of Low-Alloy Steel S/73 1569 FSAR 1.9.85 Materia l Code Case Accept. ASME lit, RG 1.85, Rev. 20, Design Div. I 11/i82 1578 FSAR 1.9.116 QA Requi rements for Installation RG 1.116, Rev. O, Construction inspection & Testing of Mechanical 6/76 Equipment and System FSAR 1.9.11 QA Requi rements for Insta l lation, ANSI N45.2.8-1875 Construction Inspection & Testing of Mechanical Equipment & Systems 1611 FSAR 1.9.139 Guidance for RHR RG 1.139, Rev. O, Design 5/78 1727 FSAR 3. NSSS Components Design Classification ANSI N18.2-1973 Design 1740 FSAR 3. Conta inment Atmosphere Cleanup RG Design 1754 FSAR 3. VEGP Classification System RG 1.29 Design
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TABLE 1 3
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COMMITMENT VERIFICATION ,,.
,
Re fe rence Commitment Commitment . . ..
Responsible Number Sou rce Section . Commitment Subiect Document ' Fea tu re Section 880 FSAR! 3.2.2-1 I rincipal. Codes and Standards for P ASME lit, Class 1, 2,' Design T.3.2.2-1 3, o r MC, NF, or CS -
885 FSAR 3.2.2-2 Const. Codes /Stds. Q.G.-B for Pressure ASME-fill, Subsection.NC Design-Vcssels, Piping, Pumps, Valves, AT ' Class 2 Storage Tanks, 0-15 psig Storage .
Tanks, Supports, Meta l Cont. Comp, Core Support Str FSAR 3.2.2-2 Const. Codes /Stds. QC-A. fo'r Pressure ASME 111,. Subsection NB,- Design Vessels, Piping, Pumps, Valves, Class 1 Supports 889 FSAR 3.2.2-2 Const. Codes /Stds. Q.G.- fo r ASME'lli Subsection ND, Design'
Pressure Vessels, Piping. Pumps Class-3 4695 FSAR 3.8.3. Cont. Int. -Structures Bolts Attaching ASME lil,' Subsection NF Design'
Pressurizer Base to Steel Support frame 1785 FSAR' 3.8.3.6- Ma te ria l for Steel Linear Supports of ASME lil, Subsection NF Design RCS 4732 FSAR 3.9.N.1. Critical Damping for OBE & SSE Seismic 2% and 4% Respectively Design-Analysis 4951 FSAR 4.A. Core Exit Thermocouple Monitorin RG 1.75 - Design
. System - Cable Routing 4997 FSAR 5.2.3. Control of Welding for Ferritic ASME 111 & IX Constructio Ma te ria l 4998 FSAR 5.2.3. Prevention of ICA No Block Welding Max 350 Construction Degree Interpass Tem Exercise Weld Proce App rova l 322 FSAR 6.1. ESF Materials' Selection & Fabrication ASME lil,' NC-2160 & NC-3120 Design 328 FSAR 6.1.1. ESF Construction Materials - Weld ASME II', Pa rt C, SFA 5.1, Design Mat'l for Joining Ferritic Base 5.2, 5.5, 5.17, 5.18 & 5.20 Mat'Is 330 FSAR 6.1.1. ESF Const. Mat'Is for Joining ASME 11, Pa rt C, SFA 5.4 & Desig Austenitic .9 331 FSAR 6.1.1. ESF Const. Ma t' I s We l d Ma te ria l ASME.Ill.& IX Design Qualification
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TdBLE 1 Lo-e COMMITMENT VERIFICATION > ',.
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Reference Commitment Commitment .
Responsible Number Sou rce Section . Commitment Sub_ lect Document Fea tu re Section 332 FSAR '6.1.1. ESF Consti Mat'Is Austenitic ASME 11 Design
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Utilized in Final Heat Treated Con FSAR 6.1.1. ESF Const. Mat'l - Cold Worked Aus Limited to No Creater Than Design ,000 psi Yield Strength 2486 FSAR 6.2. Material that Can Come in Contact with Austenitic Design Recirculation Fluid 2515 FSAR 6.2. Design of Conta inment Emergency Core RG 1.82 Design Cooling System Sumpe 2512 FSAR 6.2.2.2. NPSH Available to the Containment 'RG design Spray Pumps 4674 1. C-80/08/05 Vacuum Condition Resulting in Damage ' Low Pressure Process or . Design Co r re s . to CVCS Holdup Tanks Holdup Tanks that can Contain Radioactive Material Are Designed with Features to Preclude Vacuum Conditions 4680 1. C-81/11/20 Cate-Type Valve Closure Against Dif Maximum Pressures (psi) as ' Design Co rre Pres Flow Approaches Zero (Table).
(Affected Valves will be Modified.)
4153 NRC Cues Q48 Containment Emergency Sumps Design Di f fe rent ia l Pressure - Design Corre Capability induced by 50% Blockage with ~
Deb ri s