IR 05000424/1986049
| ML20215C419 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/01/1986 |
| From: | Conlon T, Novak T, Sinule M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215C405 | List: |
| References | |
| 50-424-86-49, NUDOCS 8612150132 | |
| Download: ML20215C419 (39) | |
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i [ O Clig UNITED STATES b
NUCLEAR REGULATORY COMMISSION o
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REGION 11
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101 MARIETTA STREET.N.W.
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ATLANTA, GEORGIA 30323 -
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Report No.:
50-424/86-49 Licensee: Georgia Power Company P.O. Box 4545 Atlanta, GA 30302 Docket No.:
50-424 Construction Permit No.:
CPPR-108 Facility Name: Vogtle Unit 1 Module:
No. 20, Instrumentation and Controls Reviews Conducted: June 2, - July 25, 1986 Inspections Conducted: June 23 - July 3, 1986 and July 21 - 25, 1986 NRC Offices Participating in Inspections / Reviews:
Office of Inspection and Enforcement (IE), Bethesda, MD Office of Nuclear Reactor Regulation (NRR), Bethesda, MD Region II, Atlanta, GA Reviewer: John Thompson, NRR Inspectors: Milton D. Hunt, Reactor Inspector, Region II Thomas McElhinney, Reactor Inspector, Region II Marvin E. Kli. Consultant, Region II (EG&G Idaho, Inc.)
Apprave s2-fd 7y,,[j[6 T. 'E. Conlon, Chief (Module--All Sections)
Date Signed Plant Systems Section Division of Reactor Safety, Region II
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t-M. V. Sinkule, Chief Da te 'Si g ned Projects Sections 3C Division of Reactor Projects, Region II 0lY Y u. L r<-
/ L ff T. Novak, Deputy Director //
Dat( Signed Division of PWR Licensing A Office of Nuclear Reactor Regulation (NRR)
8612150132 861205 DR ADOCK 0500
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i SUMMARY The Readiness Review Program is being conducted at the initiative of Georgia
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Power Company (GPC) management to assure all design, construction, and operational commitments have been properly identified and implemented at the Vogtle Electric Generating Plant Unit 1.
Module 20, issued April 4, 1986, presents an assessment of the program for design and construction of the safety-related Instrumentation and Control Systems for compliance to the Final Safety Analysis Report (FSAR) and other regulatory commitments.
This evaluation was conducted to determine if the results of the program review of instrumentation and control design, fabrication, procurement, and installation presented in Module 20 represent an effective and accurate assessment of the requirements, if the requirements were properly implemented, and if the resolutions of the findings identified in Module 20 were correct.
This evaluation was performed by NRC reviewers from the Office of Nuclear Reactor Regulation (NRR), the Office of Inspection and Enforcement (IE), and Region II inspectors. The evaluation was accomplished through a detailed examination -of all sections of Module 20 by:
1.
Assuring the accuracy of the information contained in Module 20.
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Verifying the instrumentation and control commitments identified in the Module are correct and in accordance with FSAR commitments and regulatory requirements.
3.
Checking a comprehensive and representative sample of the licensee audits and the other documents reviewed by the Readiness Review Staff (RRT) along with an independent sample of documents selected by the NRC inspectors.
4.
Inspecting a representative sample of I&C components and systems currently installed in Unit 1.
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Reviewing the results of past NRC inspections at Vogtle Unit 1 that pertain to Module 20.
6.
Reviewing the Module 20 findings and the licensee's resolutions.
During this examination, the NRC reviewers and inspectors noticed that GPC management supported the Readiness Review by actively participating in developing and implementing the program. This evaluation also indicates that the licensee's program review was comprehensive and provides adequate assurance that
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i installation and inspection of the plant instrumentation and control components and systems is in accordance with NRC requirements and FSAR commitments (except for the findings which were identified by the NRC reviewers and inspectors).
These findings should be subjected to continuing review and action until closed out in order to preclude the possibility of developing safety problems.
The Nuclear Steam Supply System (NSSS) vendor supplied instrumentation and control equipment documentation and installation were not included as part of this Module. Module 16, Nuclear Steam Supply System, was to have addressed the I&C portion, but only the in-core instrumentation system was evaluated. At the time the RRT performed its evaluation, the installation of the NSSS instrumentation and control systems were not completed in various areas of the containment and auxiliary buildings.
At the time the NRC inspectors evaluated Module 20, the NSSS I&C installation was complete. Thus, the sample was enlarged to include the NSSS I&C equipment and systems.
The findings identified during this evaluation are summarized below:
o Unresolved Item - Review 1500 psi Pressure Ratings for LT-459, LT-460, and LT-461 (URI 424/86-61-06).
o Inspector Followup Item - Examine Separation Requirements for PT-405, PT-457, and LT-462 Sensing Lines (IFI 424/86-61-07).
Inspector Followup Item - PI-0977 Is Not Installed According to ISO Drawings o
(IFI 424/86-61-08).
It does not appear that the foregoing items represent significant programmatic weaknesses. This conclusion is made with the provision that the foregoing open items for Vogtle Unit I can be satisfactorily closed. Resolution of all matters concerning these open items will be handled during future Region II inspections.
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V0GTLE ELECTRIC GENERATING PLANT UNIT 1 READINESS REVIEW PROGRAM MODULE 20-INSTRUMENTATION AND CONTROLS 1.
Scope of Review This review consisted of an examination of each section of the Module and was performed by reviewers from the Office of Nuclear Reactor Regulation (NRR) and the Office of Inspection and Enforcement (IE) along with inspectors from Region II. The Region II review was assisted by one employee of EG&G Idaho, Inc., a prime contractor to the U.S. Department of Energy at the Idaho National Engineering Laboratory. Module Sections 1.0, 2.0, 4.0, 5.0, and 8.0 presented data on Module organization, project organization, program description, audits and special investigations, and conclusions.
These did not require the review depth given to Module Sections 3.0 and 6.0 which covered Commitments and Program Verification.
The review of Module Section 7.0, Independent Design Review, is the subject of a separate report.
Sections 3.0 and 6.0 provide the more significant aspects concerning licensee commitments along with adequacy of commitment carry-through into both program implementation and design execution. Review of these two sections included an examination of content; review of findings, concerns and observations; review of a sample of items reviewed by the Georgia Power Company (GPC) Readiness Review Team (RRT); and an examination of an independently selected samples of records and field construction.
The methodology used and an evaluation of each section are i
presented in the following sections.
2.
Methodology a.
NRR Review
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The review and evaluation by the Office of Nuclear Reactor Regulation focused on the adequacy and accuracy of the commitments contained in Section 3.0 of the. Module.
This involved reviewing the Module for commitment involvement and making a detailed examination of Module Subsections 3.4 (Commitment Matrix) and 3.5 (Implementation Matrix).
The objective was to determine the extent that the licensee complied with licensing commitments for safety-related instrumentation and controls (I&C) and to determine whether all listed commitments were properly within the scope of Module 20. The primary review criteria
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for the foregoing included the Vogtle Unit 1 Final Safety Analysis Report (FSAR). Other criteria used included NRC Regulatory Guides and
related NRC staff positions.
l b.
IE Review The review and evaluation by the Office of Inspection and Enforcement focused on the Independent Design Review (IDR) Report which was not included in Section 7.0 of the Module. The results of the IE review of
this report will be the subject of a separate report.
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c.
Region II Review Review and evaluation by the Region II Evaluation Team was begun by reviewing the Module in its entirety in the offices of the team members during June 1986. The total Module was reviewed for organization and content at that time.
The first inspection was made at Vogtle Unit I during June 23 - July 3, 1986. The following activities were conducted, and the findings were documented in Inspection Report 50-424/86-61:
o Determined the RRT organization element responsible for Module 20 and interviewed key staff members o
Verified Module 20 review boundary
o Reviewed the material presented in Module 20, Sections 2.0 through 5.0 o
Obtained supplemental documentation copies required for review
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use.
In discussing the Module 20 boundary with the RRT members, it was l
1 earned that the seismically qualified I&C panels should have been i
inspected in Module 6, " Electrical Control Panels," but the I&C panels were not. Also, the Nuclear Steam Supply System (NSSS) instrumentation installation was not complete at the time Module 20 was prepared and was not included in Module 20.
Further, because only 10 commitment j
implementations out of 346 were selected for verification and only 15 instruments were selected for walkdown, the NRC reviewers chose to
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i select a relatively-larger sample for verification (27 different commitment implementations and 10 different instruments) different from those chosen by the RRT.
The Module 20 Commitment Matrix contained 346 commitments pertaining to I&C. One hundred and forty-six commitments were selected as the NRC l-sample for first-order document verification to provide the basis for initially verifying the data reported in Sections 3.0, 4.0, and 6.0 of
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the Module and to spot areas or items potentially needing review I
emphasis.
The commitments chosen were basically from the FSAR, Chapter 7.1, " Instrumentation and Controls."
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Data were gathered during the first inspection trip for office review
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prior to the second inspection trip.
The sample commitments were traced backwards into source documentation which was typically the Final Safety Analysis Report (FSAR). This was to check for proper RRT i
l recognition of the actual commitments. Most of the first inspection trip was spent in conducting walkdowns on 10 additional instruments and verifying the walkdown of 2 instruments from the RRT sample.
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The second inspection at Vogtle Unit I was made during July 21-25, 1986.
This inspection involved the following activities and is documented as part of this report o
Continuing general Module review activities o
Completing commitment tracing o
Performing Design Program Verification review o
Performing Construction Program Verification review o
Verifying Module 20 findings and licensee's responses.
The sample commitments were traced into the first-order implementation documents (design criteria and procedures) and into second-order implementation documents (drawings, specifications, and vendor submittals).
3.
Evaluations The evaluation of each Module section is provided in the remainder of this report using a Module section-by-section format. This evaluation includes a description of the section, what was reviewed, the basis for acceptance, and a statement of any required followup or evaluation, a.
Section 1.0 - Introduction (1) Review Introduction and Section Examination This section of Module 20 provides a description of the intent and content of Module 20.
Also, it provides a description of the Vogtle Unit 1 I&C hardware covered within the module, an overview of the project status, and an outline of the module organization.
This section was examined by the inspectors for content, background, and accuracy of information.
The scope was determined to cover safety-related instrumentation and controls (I&C) for the reactor coolant system, steam systems, and containment and auxiliary systems.
The RRT members who prepared Module 20 were not available for consultation with the NRC inspectors.
(a) Boundary Definition.
The information given in Module Table 1.1-1 was reviewed in detail with the RRT member assigned to Module 20 to assist the NRC reviewers.
The information gained during the review did not disclose verification erro _
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(b) Module Organization.
The Module organization portion of the section was examined by the inspectors, and no instance of inaccuracy or need for clarification was necessary.
The coordinating RRT member was asked about significant changes subsequent to the January 1,1986, cutoff date for Module 20 data.
No significant changes have occurred since that date except that the NSSS I&C installation has been completed.
No evidence of other significant module-basis change since January 1, 1986, was discovered during the review.
(c) Project Status.
The status shown in Subsection 1.3 of the Module is 96% for Design, while Construction shows 85% for instrument installation and 36% for tubing installation, as of January 1,1986.
(2)
Inspection Results The clarifications provided by the RRT, as noted above, correlated with other information reviewed by the inspectors.
The examination did not disclose significant verification errors or a basis for programmatic concern. Followup or additional evaluation of Module Section 1.0 is not required.
b.
Section 2.0 - Organization and Division of Responsibility (1) Review Introduction and Section Examination This section of the module provides a description of the organizations employed for project activities including design, field construction, procurement, quality assurance, training, and certification programs.
This section was examined by the inspectors for content and background information.
The information r-esented agrees with that obtained by the NRC l
inspectors c ring past inspections at Vogtle Unit 1.
No instances j
of variance from the Section 2.0 information were found during the course of the total module review.
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The Training and Qualification Subsection was read for content and l
general conformance with the other information contained in the l
Module. A detailed examination of the subsection was not made by
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the inspectors since the material contained was largely descriptive and not in the nature of an assessment.
The subsection describes the project programs for training and qualification of design engineers, construction engineers, contractor staff, construction crafts, and QC inspectors.
Discussion of the material contained in the Module with the RRT did not disclose information different from that presented in the module or that gained by the inspectors during past inspections.
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Inspection Results The examination did not disclose significant errors or a basis for programmatic concern.
Followup or additional review is not required.
c.
Section 3.0 - Commitments (1) Review Introduction and Section Examination This section of Module 20 describes the commitment selection and sources along with containing a list of commitments and implementing documents which are displayed in two matrices. The first is entitled " Commitment" and lists 346 commitments by the Georgia Power Company for Vogtle Unit 1 along with the source document reference for each commitment. The second is entitled
" Implementation" and lists source documents and requirement features referred to within each commitment along with the document reference where the feature has been implemented. The NRR identification review was directed at assuring that all required technical regulatory requirements relating to safety-related I&C had been included in the Module 20 Commitment Matrix listed in Section 3.4.
The Region II review was directed at verifying the proper implementation of the listed commitments.
This latter was accomplished by selecting 27 commitments as a sample.
The sample was examined by carefully checking the commitment source (typically the FSAR) for the exact requirement and verifying (within the documentation listed in the Implementation Matrix) that the requirement was carried through.
(a) Commitment Identification Review.
The NRR review of the Vogtle Readiness Review Module 20, Instrumentation and Controls, was initiated by a review of the Commitment Matrix (Section 3.4).
The review consisted of a comparison of the commitments listed in Section 3.4 and with the guidance contained in the Standard Review Plan (SRP) and other regulatory documents (SER, FSAR and Generic Letters.
The staff also reviewed the commitments to ensure that all commitments necessitated by the Standard Review Plan were included in the Readiness Review Program (all modules).
Since no questions or clarifications were found, an evalua-tion was performed based on the results of the initial review.
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Each commitment was examined to determine if it properly reflected the source (FSAR or other documents) requirement and was properly within the scope of Module 20. Also, the sources were examined to assure that all commitments were properly reflected in the Commitment Matrix.
(b) Implementation Review.
The Region II examination of Section 3.0 started with reviewing it for content.
One hundred and forty-six commitments (see Table 1) were selected as a commitment review sample for FSAR verification.
Twenty-seven commitments (see Table 2) were selected as a commitment implementation review sample. The examination of this sample consisted of the following items:
o Veri fying correspondence between the Subsection 3.4 Commitment Matrix and the Subsection 3.5 Implementation Matrix for each commitment, o
Reviewing the referenced commitment source-documentation for a clear statement of requirement for each commitment within the sample.
o Checking the document listed in the Subsection 3.5 Implementation Matrix for proper first order implementation of the requirements to meet the commitment (see Table 2).
The individual commitments reviewed along with the review results are listed in Tables 1 and 2 of this report. The anomalies that were discovered are outlined in the footnotes to the tables.
Three anomalies resulted from this inspection:
o In Section 3.4, FSAR 7.3.3.1.2 Commitment 4642 erroneously lists IEEE 279-1971, Section 4.12; should be Section 4.17.
o In Section 3.5, Commitments 1491, 2405, and 2406 do not apply to this module and should be disregarded.
o In Section 3.5, Commitment 2398 lists DC-1010, Rev. 4, as the current version of the design document rather than Rev. 5.
(2) Inspection Results The staff has reviewed Section 3.4 of Module 20 and has determined that the commitment matrix contains no omissions and is within the defined scope for Module 20.
The NRR staff has determined that the Readiness Review commitments are consistent with the Vogtle
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licensing commitments and are within the defined scope of Module 20 and are, therefore, acceptable.
The implementation review by Region II did not disclose signifi-cant implementation omissions.
d.
Section 4.0 - Program Description (1) Review Indroduction and Section Examination This section of Module 20 describes the work processes and control for design and construction of I&C systems; materials management; and fabrication, installation, and inspection.
The section was examined by the NRR inspectors for content, background, and accuracy.
(a) Design.
Subsection 4.1, entitled " Design," provides an overview of the design process and control discussed in more detail within subordinate paragraphs.
The process for preparation, control, review, and change control of safety-related I&C design documents were examined.
Design criteria provide the bases from which the I&C design is controlled.
The Regulatory Guides, codes, standards, calculations, and operating plant experiences were used to determine I&C requirements.
Drawings, specifications, data sheets, logic diagrams, loop diagrams, and reports were prepared to facilitate construction, procurement, and licensing activities, as well as subordinate engineering design functions. These documents contain the criteria that govern the-design of safety-related I&C.
Design criteria were initially provided in the Preliminary Safety Analysis Report (PSAR). Detailed design criteria were subsequently developed and issued as part of the Vogtle Project Design Manual.
Design verification reviews were systematically performed to ensure that equipment, systems, and structures were properly designed and that the designs were properly coordinated.
Design changes for I&C documents and drawings were achieved I
by issuing Design Change Notices (DCNs) or Interim Design Change Notices (IDCNs) or by Field Change Requests (FCRs).
Deviation Reports (DRs) were used to report deficiencies in material, documentation, or installation procedures, identified during the construction phase of the project.
General agreement was found among the Module description, procedures, and observed practices for all examined documents that related to Subsection 4.1.
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(b) Materials.
Module 20, Subsection 4.2, refers to Module 21 Appendices C and E, entitled " Procurement" and " Material Control," respectively.
A detailed examination of the subject of those appendices was not performed due to the absence of the material from Module 20.
The Materials Subsection was reviewed for content and conformance to commitments, however.
In addition, the commitment sampling review listed for Module 20, Section 3.0, was carried into the corresponding procurement specifications in several' cases to verify commitment implementation. General agreement among the material program description, commitments, and implementation was found for all materials investigated under Module 20, Subsection 4.2.
Georgia Power Company (GPC) purchases, receives, inspects, and stores instruments, equipment, and materials, including welding material required for construction at Vogtle Plant.
Material required for construction is controlled in accordance with the general site receipt inspection program, which covers receipt, receipt inspection, document review and acceptance, release for issue, warehouse storage, and in place storage.
(c)
Fabrication, Installation, Inspection, and Testing.
Safety-related I&C equipment is procured or fabricated from materials procured in compliance with specifications which are prepared in accordance with the design criteria contained in the Vogtle Electric Generating Plant (VEGP) Project Design Manual. Detailed descriptions of the procurement process and receiot and document review are contained in Appendices C and E,
respectively.
Vendor requests for approval to deviate from the requirements of an equipment or material specification were submitted on a Supplier Deviation Disposition Request (SDDR) Form.
i Subsection 4.4 was examined by the inspectors for content and general agreement with the Module 20 commitments selected for review.
In addition, the flow charts contained in this subsection were reviewed for general logic and accuracy.
General agreement was found to occur between the commitments and activities covered by the Fabrication, Installation, and Inspection Program.
(2) Inspection Results The Section 4.0 examination did not disclose verification errors or further basis for programmatic concerns.
Followup or additional evaluation is not required.
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e.
Section 5.0 - Audits and Special Investigations (1) Review Introduction and Section Examination This section provides a discussion of the audits of the Module 20 related items made by GPC and Bechtel Quality Assurance organizations, along with those performed by NRC and the Licensee's Self-Initiated Evaluation Team.
It also included a discussion of past design and construction problems considered by GPC to be important and potentially reportable to the NRC under 10 CFR 50.55(e). This section was examined by the1NRC inspectors to confirm that the process for reportability evaluation was used.
Georgia Power Company conducted regularly scheduled audits to verify compliance with project requirements.
Bechtel Power Corporation conducted scheduled audits to verify compliance with project procedures, national codes and standards, and NRC Regulatory Guides. The NRC conducted several inspections relating to I&C installation.
The Self-Initiated Evaluation (SIE) Program was conducted for an in-depth review of design and construction activities, using criteria developed by the Institute of Nuclear Power Operations (INPO).
Several project audits were conducted that represent broad, in-depth reviews of construction activities related to I&C installation. These audits principally addressed work activities as they progressed, including control of design documents, installation activities, and quality assurance (QA) documentation.
Seven problems related to I&C design were identified during the course of VEGP design. In each case, the NRC was informed of the potential reportability of the deficiency and was subsequently informed that the identified deficiencies constituted a reportable condition pursuant to the criteria of 10 CFR 21 and 10 CFR 50.55(e).
(2) Design Program Audits Bechtel Power Corporation conducted seven design audits. Georgia Power Corporation conducted two design audits, and NRC conducted one design audit.
Audit GD11-81/61 resulted in Findings 250 and 251.
Finding 250 concerned preliminary drawings having been issued in Series CX5DP002 and CX5DP005.
Bechtel was requested to issue completed, signed-off drawings for construction; and the finding was satisfactorily closed.
Finding 251 concerned Nonconformance Reports that had not been approved by the design organizatio * -
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Approval was obtained, and the finding was satisfactorily closed.
The Module states in Paragraph 5.1.1 that no design-related I&C findings were identified in NRC Inspection Report 50-424/425/83-19. However, the Design Audit Table (at the end of Module Section 5.1) shows one finding for this audit as follows.
" Unresolved Item 50-424/425/83-19-01, ' Determine a Method for Incorporating CDR Resolutions in Design Document'.
The NRC inspector determined that there appeared to be na set procedure or method to ensure that changes made in equipment locations or changes made in plant design as a result of CDR resolutions were followed up by the licensee to ensure incorporation of the change in the necessary design documents. The licensee's representatives advised the inspector that follow up was taking place, but the method used was not addressed in a procedure.
This will be identified as an unresolved item."
The eight BPC audits resulted in seven findings, all minor in nature, and corrective actions were completed promptly.
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Following submittal of Module 20, BPC identified several Bechtel Power Corporation audits / findings that had been left out of Section 5.
These audits and associated findings were supplied to the NRC reviewers during the second site visit and have been reviewed. BPC feels that the additional findings contained in the additional audits would not have modified the verification plan due to their insignificant nature.
These additional audits are listed in the attached Table 3.
The Self-Initiated Evaluation (SIE) Program conducted an in-depth review of design and construction activities and identified one design-related finding, Finding DC 2-08.
This finding concerned failure to establish a setpoint document to identi fy/establi sh instrument setpoints.
This finding was satisfactorily resolved when such a document was developed and issued.
The setpoint document was included in the design program verification activities as presented in Section 6.1.
(3) Construction Audits Bechtel Power Corporation conducted one field construction audit.
Georgia Power Company conducted fifteen QA audits. NRC conducted nine on-site inspections.
The 15 GPC audits resulted in 27 findings or deficiencies related to construction activities.
The RRT concluded that all findings l
were of moderate importance. The NRC inspectors concurred with the RRT conclusions.
The nine NRC inspection reports identified two Severity Level IV violations. GPC has submitted responses to these two violations which are being reviewed by the NR.
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(4) Past Design and Construction Problems Section 5'.3.1, Module 20, states that seven design problems were identified; however, only six problems were addressed. It appears that six problems is the correct number. These six problems are discussed below.
Six problems related to instrumentation and controls (I&C) design
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were identified. during the course of-VEGP design. In each case, the,NRC was informed of the potential reportability of the deficiency and ' was subsequently informed that the identified deficiencies constituted a reportable condition pursuant to the criteria of 10 CFR 21 and 10 CFR 50.55(e). These six problems are described below:
(a) Hydrogen Recombiner Power Supply - Non-Class IE Power
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Supp]y - Problem involved a non-Class IE transformer that was employed in : the hydrogen recombiner power supply unit.. To resolve this concern, the hydrogen recombiner was returned to Westinghouse, the subvendor, for replacement of the transformer.
(b) Reactor Coolant System Wide-Range Pressure Transmitter
, Inaccuracy - Problem involved possible inaccuracies of the
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' reactor coolant system (RCS) wide-range pressure transmitters if used during post-accident applications.
To resolve this
- concern, two pressure-sensing lines from the RCS hot legs were established.
Each line contains two transmitters, having an increased range, located outside containment. The four transmitters provide indication on the main control board in the control roem.
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(c) ' Undetectable Failure-Solid-State Protection System
,The problem involved a potentially undetectable failure in e
on-line testing circuits for relays in the Solid-State Protection System. The condition was resolved by corrective action to testing procedures to ensure that no failures would occur in the circuits after ermpletion of testing.
(d) Defective Printed Cird Cv 1s -
The problem involved defective heat sink a ab'
s on printed circuit cards for the nuclear loop pow +.. sum. The condition was resolved by returning the defective circuit cards to the vendor for repair.
(ei Isolation Valves and Damper Actuators-Defective Seals - The problem involved rubber. seals used in rotary actuators for valves and dampers that ' swell when in contact with Mobil 28 lubricant or flatten if net exercised.
Four valves and a s
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number of dampers were identified to be affected by the defective actuators.
To resolve the deficiency, each installed actuator was repaired with new seals and lubricant.
Deficient actuators in storage have been identified and will be serviced with the new seals and lubricant prior to installation.
(f) Undetectable failure-Engineered Safety Features Actuation-System - The problem concerned an undetectable failure involving an interlock used to input the status (open or closed) of the reactor trip breakers to the Engineered Safety Features Actuation System. To resolve the condition, the system testing procedure was revised to identify this type of failure as prescribed by IEEE 379, "Applicat'on of Single Failure Criterion to Nuclear Power Generating Station Protection Systems."
(5)
Inspection Results The twenty additional BPC audits were reviewed by the NRC inspectors, and no significant discrepancies or findings were noted.
No significant inaccuracies were found. Review of the past design
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and construction problems listed in Subsection 5.5 of the Module did not disclose information differing from that previously obtained by the inspectors or indicate incorrect evaluation by the RRT.
The examination did not disclose significant verification errors or a basis for programmatic concern.
Followup or additional evaluation of Section 5.0 is not required.
f.
Section 6.0 - Program Verification This section of Module 20 describes the activities undertaken (1) to ascertain whether the design and construction work processes have been adequately controlled in order to ensure implementation of licensing commitments, and (2) that the results of these work processes conform to project procedures and designrequirements. The section is further divided into subsections covering Design Program Verification and Construction Program Verification.
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(1) Design Program Verification The Design Program Verification Subsection of the Module focused on the programmatic aspects of design with the objective of determining whether the design control process functioned effectively and whether it ensured proper implementation of
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licensing commitments.
Verification was performed in two phases.
Phase I consisted of verifying implementation of all technical commitments that were within the scope of the Module.
The verification included instruments selected from three systems:
primary coolant system, main steam system, and component cooling water system. These were reviewed for proper implementation into the project design criteria and the procedures which were referred to as first-order design documents. A selected portion of the
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commitments were reviewed further for implementation into specific second-order design documents including calculations, drawings, specifications, and vendor submittals.
Phase II consisted of reviewing selected second-order design documents for compliance with project procedures and industry standards.
(a) Phase I Examination. The Phase I part of the Design Program Verification Examination started first with an NRC Region II inspector's selection of a sample of 146 commitments from the 346 listed in Subsection 3.4 of the Module.
This sampling included 7 of the 27 being identified for design cognizance within the Module 20, Subsection 3.4 Matrix. The commitment sample and the results of the review are listed in Table 1 of this report.
Two commitments shown in Table 6.1-2 of the Module were examined also. Examination details are presented below.
1.
First Order Verification.
Verification of commitments was found in Module-listed first order documents for the 24-commitment sample (see Table 1). In reviewing Module Section 3.4, Commitment 4642 (FSAR 7.3.1.2.2.7) has a discrepancy in that Section 4.12 is listed erroneously and should be Section 4,17.
2.
Second Order Verification.
The second part of the examination of Part I involved commitment verification in second-order documents. The NRC inspector reviewed 2 of the 10 commitments listed by the RRT in Module 20 Table 6.1-2.
The results of the NRC examination are listed in Table 2 of this report.
In addition, the Pullman Power Products (PPP) construction packages and the GPC procurement packages for the 12 instruments selected for walkdown (see Table 4) were reviewed.
These packages typically contained design specifications, purchase orders, equipment / material receipt inspection reports, field change notices, walkdown verifications, maintenance work orders, deviation reports, certified materials test reports, liquid penetrant examination records, QC Inspection check lists, and rework / inspection forms. No cases of significant second-order document verification-error l
were found in the commitments checked by the NRC inspectors.
I
l
. _
._
--
- - - _
_ _.-.
-
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.
..
.
,
.
Several anomalies were found. Commitments 1491, 2405, and 2406 were found to be not applicable to Module 20.
In Table 6.1-2 of the Module, Commitment 4641 (FSAR 7.3.1.2.2.7)
should be 4642, and for Commitment 4557 (FSAR 7.6.7.1)
the criteria design section should be DC-1201 instead of DC-1701.
(b) Part II Examination.
Part II of the RRT Design Program Verification involved a detailed review of samples taken from the following design categories:
o Field Change Requests o
Deviation Reports o
Field Walkdowns o
Specifications o
Findings.
All of the foregoing module topics were examined.
It was noted that the RRT used predeveloped checklists to review each of the foregoing documentation. categories.
no verification errors were found.
(2) Construction Program Verification (a) Review Introduction and Subsection Examination.
The Con-struction Program Verification Subsection of the Module focused on the programmatic aspects of construction.
The objective was to determine whether the construction control process functioned effectively and whether it ensured proper implementation of licensing commitments.
The RRT verifi-cation was performed in three phases.
Part I consisted of verifying the implementation of all construction commitments that were within the scope of the Module.
These were reviewed for proper implementation in the appropriate construction process documentation. Part II was a technical review of the construction records to ascertain whether the necessary documentation requirements were verifiable and to establish confidence in the quality of the work done.
Part III consisted of a field walkdown of selected hardware items and components to verify that each was in conformance with the appropriate design drawings, construction specifications, and installation procedures.
The PPP construction packages were examined for the 12 instruments selected for walkdown (see Table 4).
No deficiencies were noted.
(b) Part III Examination.
Part III of the RRT Construction Program Verification involved a detailed walkdown review of the I&C categories. Twelve instruments (10 of them different from the RRT sample) were walked down by the NRC inspectors
.
-
.
-
-
-
_.
- -. -.
- - -. _ _.
..
,
.
.
,
(see Table 4).
This was done to ascertain whether the installation was in accordance with the design drawings, construction specifications, and procedures.
The examination of the construction Program Verification Subsection resulted in several minor anomalies which are discussed below:
o Several instances of instrument sensing lines having incorrect slope were noted (PT-1853, PI-1879, PT-0601, and PI-0977).
The response was that this would be corrected on the final QC walkdown. The response given by RRT personnel was acceptable to the NRC inspectors.
o One damaged pressure indicator was noted-(PI-402 had a badly bent setpoint indicator and the indicating needle was pegged out upscale).
The response was that this would be corrected during functional checkout by Operations.
The response given by RRT personnel was acceptable to the NRC inspectors.
o Sensing tubing holddown clamp is missing on PI-17597.
The response was that this would be noted during QC final walkdown and fixed.
The response given by RRT personnel was acceptable to the NRC inspectors.
o Sensing tubing touches flexible conduit on PT-405. The response was that this would be corrected by means of a Maintenance Work Order.
The response given by RRT personnel was acceptable to the NRC inspectors.
o Installation drawings do not agree with actual physical installation (LT-0990 and PI-0977).
The response was that the drawings were in the process of being changed by means of a DCN. The response given by RRT personnel was acceptable to the NRC inspectors.
o Discrepancy was observed on instrument nametag data for Pressurizer Level Transmitter LT-0461 (IFI 424/86-61-06).
When the Pressurizer Level Transmitter LT-0461 was inspected, it was noted that the manufacturer's nametag (Westinghouse Veritrak)
contained the following information:
Serial number, model number, and the words " safe pressure 1500 psi."
This safe pressure wording caused concerned bertuse the pressurizer operating pressure will be a proximately 2350 psi.
r Basically, four instruments are redundantly performing this same function:
LT-0459, LT-0460,
-
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.
.
.-_
-
_
F
.
,
LT-0461, and LT-0462.
LT-0459, LT-0460, and LT-0461 nametags all contain the words " safe pressure 1500 psi;"
LT-0462 nametag is marked " maximum pressure 2500 psig."
The NRC inspectors met with RRT personnel and vendor personnel located on site.
The NRC inspectors were assured that these instruments would indeed work satisfactorily in this service.
The on-site vendor personnel had contacted their design engineers and verified the following information.
Their standard practice was to take instrument blanks marked " safe pressure 1500 psi" and upgrade them by testing to operate at higher pressures. The vendor stated that the nametags were erroneously left on the transmitter body, but test records showed the instruments had been upgraded to operate at 2500 psig. The NRC will continue to pursue this matter during subsequent inspections.
The above discrepancies have been combined into three findings which are identified in Section 4.
The subsection examination did not disclose substantial verification errors.
Follow-up or additional evaluation, other than noted above, is not required.
g.
Section 7.0 - Independent Design Review The results of the Independent Design Review (IDR) were not included in Module 20. A brief narrative of the IDR process was provided. This included a statement that an independent engineering consulting company had made a review of the design documents (such as design criteria, calculations, specifications, and drawings) to ascertain whether these documents correctly implemented licensing commitments.
A team of technical and professional experts assessed the adequacy of the design work for this Module. The Independent Design Review will be examined by the NRC Office of Inspection and Enforcement. The results of this NRC examination will be the subject of a separate report.
h.
Section 8.0 - Program Assessment / Conclusions (1) Review Introduction and Section Examination This section of Module 20 provides a summary of open corrective actions, certifications from review managers or participants, and mini-resumes for the various RRT members.
No introductory paragraph was provided to explain the significance of the information as had been given in the other sections.
The information presented and the lack of explanation of its significance relative to the section title is not in keeping with the general descriptive lead-in contained in the other sections of the module.
[-
r
5
-
The examination of this section by the NRC inspectors involved reviewing the section for content and background information.
Only one finding (20-11) was listed in the Module as being open and requiring action by BPC.
The NRC reviewers inspected the revised level setting diagrams and setpoint list and found them to be satisfactory. This finding should be considered to be closed.
(2)
Inspection Results The Section 8.0 examination did not disclose substantial verification errors or a basis for programmatic concern. Followup or additional evaluation, other than noted in the Section 6.0 instrument walkdowns, is not required.
4.
Findings The following three findings were identified during the NRC evaluation of the Module. All deficiencies noted are considered to have minimal safety significance at this point.
These have been identified as Inspector Followup Items (IFI) based on the nature of the followup action required.
These will be addressed by the NRC during the routine inspection program.
a.
Unresolved Item - Review 1500 psi Pressure Ratings for LT-459, LT-460, and LT-461 (URI424/86-61-06).
When the Pressurizer Level Transmitter LT-0461 was inspected, the manufacturer's nametag (Westinghouse Veritrak) contained the following information:
Serial number, model number, and the words " safe pressure 1500 psi."
This safe pressure wording caused the inspectors to be concerned because they knew the pressurizer operating pressure was approximately 2300 psi.
Basically, four instruments are redundantly performing this same function:
LT-0459, LT-0460, LT-0461, and LT-0462.
LT-0459 LT-0460, and LT-0461 nametags all contain the words " safe pressure 1500 psi;"
LT-0462 nametag is marked " maximum pressure 2500 psig." Are the three level transmitters marked " safe pressure 1500 psi" qualified to operate in this service?
b.
Inspector _ Followup Item - Examine Separation Requirements for PT-405, PT-457, and LT-462 Sensing Lines (IF1424/86-61-07).
Pressure Transmitter (PT-405) sensing line was observed to be touching a flexible power conduit containing 117 volts, 60 hertz, lighting circuits.
This violates the Vogtle Program Design Manual.
c.
Inspector Followup Item - PI-0977 Is Not Installed According to ISO Drawings (IFI424/86-61-08). The installation drawings do not agree with the actual physical installation.
5.
Conclusions The NRC has reached the following conclusions for instrumentation and controls at Vogtle Unit 1 based on the review of Module 2 p
.
.
a.
Summary of Specific Conclusions The following Module sections have been determined to be acceptable with the exception of items and areas discussed earlier in this report.
(1) Section 1.0 - Introduction--The boundary between Module 20 and the related Modules is generally clear and well defined as presented in Section 1.0.
Minor clarification of the data presented in Table 1.1-1 was required for definition completeness. The Module Organization and Project Status were correct as of the date of Module publication.
(2) Section 2.0 - Organization and Division of Responsibilities -- The organization description and responsibility presented in Section 2.0 of the Module were reviewed and verified as being correct.
(3) Section 3.0 - Commitments--The commitments listed in Section 3.0 were reviewed and determined to be complete and correctly identified (to be confirmed by NRR).
Implementation was verified in first-order documents.
(4) Section 4.0 - Program Description--The design program description presented in Section 4.0 was verified as being correct.
The information presented concerning materials, and construction was verified as being correct.
(5) Section 5.0 - Audits and Special Investigations--The audits and special investigations information presented in Section 5.0 was reviewed and verified as being correct. Several additional audits (see Table 3) were given to the NRC reviewers during the second site visit and were reviewed for content. No significant findings were identified.
(6) Section 6.0 - Program Verification--The design program verifica-tion reported in Subsection 6.1 of the Module was verified as being generally adequate.
The examination of a wide range of documents provided verification of adequate documentation-system and programmatic functioning. The examination of the RRT walkdown, and the 10 additional instrument walkdowns, did not result in verification errors. The design program verification performed by the RRT is censidered to be sufficient to provide a conclusion of adequate programmatic functioning.
The construction program verification reported in Subsection 6.2 was verified as being adequate.
The instrument installations that were examined reflected adequate construction practices.
As a result of the examinations conducted as part of Section 6, two unresolved items and one inspector followup item were identified. These items are discussed in Section e
"
e
(7) Section 7.0 - Independent Design Review--The licensee engaged the Stone and Webster Company to perform an independent review of the design. This review as made in conjunction with the design review of other related Modules. The IDR is documented in Module 22 and its associated report.
(8).-Section 8.0 - Program Assessments / Conclusions--The summary of corrective actions presented in Section 8.0 of the Module was examined and the current status was determined.
The certifications and mini resumes did not conflict with information contained elsewhere in the Module.
The section lacked an introductory statement to define the significance of the information that it contained. An examination of the status of the one open item (Finding 20-11) disclosed that appropriate action had been taken.
b.
General Conclusions The examination performed by the NRC indicated that GPC management supported the Readiness Review by active participation and adequate resources.
No evidence of coercion or attempt to dilute either the effort or the findings was disclosed. The RRT displayed the requisite competence and professionalism for a review of this nature.
The licensee's program was comprehensive and provided adequate assurance that the plant instrumentation and controls is in accordance with NRC requirements and FSAR commitments.
The NRC concludes that the GPC program for instrumentation and controls complies with NRC requirements and FSAR commitments. This conclusion is based on information currently available to the inspectors and reviewers.
Should subsequent contradictory information become available, it will be evaluated to determine its effect on the above conclusion.
6.
References a.
Vogtle Electric Generating Plant, Readiness Review, Module 20, Instrumentation and Controls.
b.
Letter to NRC, from D. O. Foster, Vice President and Project General Manager, Vogtle Project, Georgia Power Company, dated April 4, 1986, forwarding Module 20 to NRC for evaluation.
1
i TABLE 1.
COMMITMENTS Numb 3r Source Section Subjtct Document /FsStu ro Rifo rGnca FSAR 1.9.84 Design and fab. code case accept.
RG 1.84, Rev. 20 1565 ASME lit, Div. 1 FSAR 3.9.B.1.5.1 Wold inspection acceptance criteria ASME Code, Section lit, 1977 5026 Edition through Winter 1977 Addendum FSAR 3.9.B.1.5.1 Weld inspection acceptance criteria ANSI B31.1 Code, 1977 Edition through 5027 Winter 1977 Addendum FSAR 3.9.B.1.5.2 Weld inspection acceptance criteria AWS D1.1-1975 5028 FSAR 7.1 I &C i n t ro.
IEEE 279-1971 4359 FSAR 7.1.2.1.1A identification of safety criteria Min. DNBR shall not be less than 1.30 4478 reactor trip system as a result of any anticipated transient or malfunction (Condition 11 events)
FSAR 7.1.2.1.28 Engineered safety features actuation The ESFAS must have provisions in the 4479 system--manual act, requirements control room for manua l ly initiating the functions of the ESF systems FSAR 7.1.2.1.3A Instrumentation and control power Vital ac the inverter sha ll have the 4480 supply system--RTS capacity and regulation requi red for the ac output for proper operation of the equipment supplied FSAR 7.1.2.1.3B Instrumentation and control power Vital ac redundant loads shall be 4481 supply system--RTS assigned to dif ferent distribution pa ne l s wh ich a re supp l i ed f rom d i f fe rent inverters FSAR 7.1.2.1.5 Interlocks (protective system)--RTS IEEE 279-1971, Sec. 4.11-4.13 4576 FSAR 7.1.2.1.5 Interlocks ( protective system)--RTS 10 CFR 50, App. A, CDC 20 4577 FSAR 7.1.2.1.5 Interlocks (protective system)--RTS 10 CFR 50, App. A GDC 21 4578 FSAR 7.1.2.1.5 Interlocks (protective systems)--RTS 10 CFR 50, App A, CDC 22 4579
.
.
TABLE 1.
COMMITMENTS (continued)
Number Source Section Subject Document /Fca tu ra Raf a r;nc2 FSAR 7.1.2.1.6 Design bases bypasses--RTS lEEE 279-1971, Sic 4.11-4.14 4580 FSAR 7.1.2.1.7 Design bases equipment protection--
IEEE 384-1974 4581 RTS FSAR 7.1.2.1.7 Design bases equipment protection--
RG 1.75 4582 RTS FSAR 7.1.2.1.8 Design bases diversity--RTS IEEE 279-1971 4583 FSAR 7.1.2.1.3 Instrumentation and control power Vital ac no single failure shall 4803 supply system--RTS cause a loss or power supply to more than one distribution panel FSAR 7.1.2.1.7 Design bases: equipment protect ion--
IEEE 279-1971 4804 RTS FSAR 7.1.2.1.9 Bi-stable trip setpoints, reacto r Do not require process transmitters 4585 protection and ESF system--RTS to operate within 5% of the high and low end of their calibrated span or ra ng e.
FSAR 7.1.2.2 Independence of redundant safety-10CFR 50, APP. A, GDC 22 4487 related systems reactor protection FSAR 7.1.2.2 Ceno ra l physical sepa ra tion criteria RG 1.75 4488 FSAR 7.1.2.2.1A Gene ra l NSSS equipment IEEE 384-1974 4489 FSAR 7.1.2.2.1B Gene ra l : Westinghouse class 1E RG 1.75, position C.4 4586 Protection system FSAR 7.1.2.2.18 Gene ra l : Westinghouse class 1E IEEE 279-1971 4587 protection system FSAR 7.1.2.2.1B Cene ra l : Westinghouse class 1E WCAP-8892-A, June '77 4588 protection system FSAR 7.1.2.2.1C Gene ra l : Physical sepa ration criteria IEEE 384-1974, Sect. 5.7 4589 Westinghouse NSSS FSAR 7.1.2.2.2 I&C Specific systems manual reacto r RG 1.62 4590 trip fcn, man, est act.
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TABLE 1.
COMMITMENTS (continuid)
.
.
Numb 3r Source Section Subject Document /Featu ro R3fer nca -
FSAR 7.1.2.2 Independence of, redundant safaty +
-lEEE.279-1971, Sect. 4.6 4805 related systems reactor protection
~
i FSAR 7.1.2.6 Westinghouse protection system and lEEE 379-1972 4593
!-
l FSAR 7.1.2.6 Westinghouse protection system and RG 1.53 5944 j
eBOP ESFAS design
\\
FSAR 7.1.2.6 Westinghouse protection system and IEEE 279-1971, Sect. 4.2 4595
'
,
j FSAR 7.2.1.1.1A Reactor trip system function Auto-initiates whenever necessa ry 4490 performance requi rements to prevent fuel damage for an j
anticipated oper. t ra n s i en t
. condition II)
i FSAR 7.2.1.1.1B Reactor trip system functional Auto initiates to limit core damage 4491 performance requi rements for infrequent faults (condition til)
,
}
FSAR 7.2.1.1.1C Reactor trip system functional Auto initiates so that the energy 4492 i
performance requirements gen, in the core is compatible with I
the design prov. to protect the j
rope for limiting fault cond.
'
(condition IV)
i FSAR 7.2.1.1.2F Reactor trip systems reactor trip on TMl action item ll K 3.10 4493
~
turbine trip i'
FSAR 7.2.1.1.2F Reactor trip systems IEEE 279-1971 4494
reactor trip on turbine trip i
FSAR 7.2.1.1.2H Reactor trip system RG 1.62 4495
,
FSAR 7.2.1.1.11 Reactor trip system 10CFR50, App. A GDC 2 4496
seismic design i
FSAR 7.2.1.2 Reactor trip system IEEE 279-1971, Sect. 3-4596 i
design bases info.
'
FSAR 7.2.1.2.5 Reactor trip system lEEE 279-1971 4597 j
abnorma l events
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t
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4 i
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TABLE 1.
COMMITMENTS (ctntinuid)
Numb 3r Sou rce Section SubJtct Document /Fssturs R2fo rcnca FSAR 7.2.2.2 Analyses 10CFR50, App. A, GDC 10 4598 reactor trip eva l, of design limits FSAR 7.2.2.2.1 Ana lyses reactor trip eva l. of 10CFR50, App. A, GDC 20 4599 design limits FSAR 7.2.2.2.1 Analyses reactor trip 10CFR50, App.
A, CDC 10 4600 trip setpoint discussion FSAR 7.2.2.2.1 Analyses reactor trip 10CFR50, App. A, GDC 15 4601 trip setpoint discussion FSAR 7.2.2.2.1 Analyses reactor trip 10CFR50, App. A, CDC 20 4602 trip setpoint discussion FSAR 7.2.2.2.1 Analyses reactor trip 10CFR50, App. A, GDC 29 4603 trip setpoint discussion FSAR 7.2.2.2.1 Analyses reactor trip BTP ICSB 12 4604 trip setpoint discussion FSAR 7.2.2.2.1 Ana lyses reactor trip 10CFR50, App. A, CDC 21 4605 trip setpoint discussion FSAR 7.2.2.2.3 Ana lyses reactor trip eva l. of IEEE 279-1971, Sect. 4 4606 compliance to codes and standards FSAR 7.2.2.2.3B Analyses reactor trip single 10CFR50, CDC 21 4607 failure criterion FSAR 7.2.2.2.38 Analyses reactor trip single 10CFR50, CDC 22 4608 failure criterion FSAR 7.2.2.2.38 Analyses reactor trip single 10CFR50, GDC 23 4609 fa i lure cri te rion t
FSAR 7.2.2.2.3D Analyses reactor trip equipment 10CFR50, App. A, CDC 4 4611 qualification FSAR 7.2.2.2.3F Analyses reactor trip independence 10CFR50, App, A, CDC 21 4612 FSAR 7.2.2.2.3F Ana lyses reactor trip independence 10CFR50, App. A, CDC 22 4613
.
TABLE 1.
COMMITMENTS (continued)
Number Sou rce Section Subject Document /Fea tu re RIfs rznc3 FSAR 7.2.2.2.3G Analyses reactor trip control &
10CFR50, App. A, CDC 24 4614 protection system i nte ract ion FSAR 7.E.2.2.3G Analyses reactor trip control &
IEEE 279-1971, Sect. 4.7 4615 protection system interaction FSAR 7.2.2.2.3J3 Analyses reactor trip capability for IEEE 279-1971 4616 testing solid-state logic testing FSAR 7.2.2.2.3J3 Analyses reactor trip capability for IEEE 338-1975 4617 testing solid-state logic testing FSAR 7.2.2.2.3J3 Analyses reator trip capability for 10CFR50, App. A, CDC 21 4618 testing solid-state logic testing FSAR 7.2.2.5 Analyses reactor trip tests and IEEE 338-1975 4619 inspections FSAR 7.3.1.1.1.1.1 NSSS, Analog initiating circuitry RG 1.1 4626 FSAR 7.3.1.2 BOP systems interfacing ESF WCAP 8760 4635 system FSAR 7.3.1.2.2 Compliance with standards and IEEE 279-1971 4636 design criteria FSAR 7.3.1.2.2.1 ESFAS single failure criteria 10CFR50, App. A, GDC 21 4637 FSAR 7.3.1.2.2.1 ESFAS single failure criteria 10CFR50, App. A GDC 23 4638 FSAR 7.3.1.2.2.1 ESFAS single failure criteria IEEE 279-1971 4639 FSAR 7.3.1.2.2.6 Manual resets and blocking features IEEE 279-1971, Sect. 4.12 4640 FSAR 7.3.1.2.2.7 Manual initiation or protective RC 1.62 4641 actions FSAR 7.3.1.2.2.7 Manual initiation of protective IEEE 279-1971, Sect. 4.17 4642 actions FSAR 7.3.3.1.2 Containment comb. gas control No single failure prevents the 4514 system containment combustible gas control system from functioning FSAR 7.3.3.2.8 Containment combustible gas control RG 1.7 4643 system
.
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TABLE 1 COMMITMENTS (continu d)
Number Source Section Subject Document / Featu rs RafGrsnca FSAR 7.3.3.2.C Containment combustible gas control IEEE 279-1971 4644 system FSAR 7.3.4.2.8 Containment purge isolation system IEEE 279-1971 4520 FSAR 7.3.7.1.1E AFWS - design bases No single failure shall prevent 4516 this system from operating FSAR 7.3.7.1.2 AFS design bases AFS - the system must provide full 4523 auxiliary feedwater flow within 1 minute FSAR 7.3.7.2.A AFS analysis of compliance 10CFR50, App. A, CDC 13 4524
FSAR 7.3.7.2.A AFS analysis of compliance 10CFR50, App. A, GDC 19 4525 FSAR 7.3.7.2.A AFS analysis of compliance 10CFR50, App. A, CDC 34 4526 FSAR 7.3.7.2.8 AFS analysis IEEE 279-1971 4527 FSAR 7.3.8.2.0 Main steam feedwater system IEEE 279-1971 4528 analysis FSAR 7.3.9.2.8 NSCW analysis IEEE 279-1971 4529 FSAR 7.3.10.2 Component cooling water system IEEE 279-1971 4498 FSAR 7.4.1.2.1J System required for safe shutdown IEEE 279-1971, Sect. 3 4645 atmospheric steam relief system FSAR 7.4.1.2.2 System required for safe shutdown 10CFR50, App. A, CDC 13 4646
]
analysis
)
FSAR 7.4.1.2.2 System required for safe shutdown 10CFR50, App. A, CDC 19 4647 analysis q
j FSAR 7.4.1.2.2.A.2 System requi red for safe shutdown 10CFR50, App. A, GDC 34 4648 analysis FSAR 7.4.1.2.2.B.2 System required for safe shutdown RG 1.29 4649 analysis
)
FSAR 7.4.1.2.2.C System required for safe shutdown IEEE 279-1971 4650 analysis
4 l
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e
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TABLE 1.
COMMl!MENTS (continued)
Number Sou rca Section Subjte t Document / Festu rs R;farenca FSAR 7.4.1.3.1.J System required for safe shutdown IEEE 279-1971, Sect. 3 4651 centrifugal charging system controls FSAR 7.4.1.3.2.A.1 System required for safe shutdown 10CFR50, App. A, CDC 13 4652 CCSC analysis FSAR 7.4.1.3.2.A.1 Systems required for safe shutdown, 10CFR50, App. A, CDC 19 4653 CCSC analysis FSAR 7.4.1.3.2.A.2 System requi red for safe shutdown 10CFR50, App.
A, GDC 34 4654 CCSC analysis FSAR 7.4.1.3.2.B.2 System required for safe shutdown RG 1.29 4655 CCSC analysis FSAR 7.4.1.3.2.3 System requi red for sa fe shutdown IEEE 279-1971 4656 CCSC analysis F3AR 7.4.3.1.3.A Safe shutdown f rom outside control Shutdown panels & safety grade 4519 room - design bases switches designed to withstand SSE with no loss of essential functions FSAR 7.4.3.1.3 Safe shutdown from outside the 10CFR50, App.
A, CDC 19 4657 control room - design bases FSAR 7.4.3.1.3.B Safe shutdown f rom outside the IEEE 279-1971 4658 cont ro l room - design bases FSAR 7.4.3.1.3.0 Safe shutdown f rom outside the Shutdown panels a re designed to 4660 control room - design bases achieve cold shutdown where offsite powe r i s ava i l a b i c a nd whe re o f f s i te powe r i s no t ava i l a b l e fo r 72H FSAR 7.4.3.1.3.A&F Safe shutdown f rom outside Shutdown panels designed to a 4665 control room - design bases coincident single failure & SSE FSAR 7.4.3.2.A Safe shutdown f rom outside the 10CFR50, App. A, GDC 19 4661 control room - analysis FSAR 7.4.3.2.B Safe shutdown from outside the RG 1,29 4662 control room - ana lysi s FSAR 7.4.3.2.8 Safe shutdown from outside the RC 1.22 4806 control room - analysis
.
O
TABLE 1.
COMMITMENTS (continued)
Number Sou rca Section SubJtct Document / Feature Rgra renca FSAR 7.4.3.2.C Safe shutdown from outside the IEEE 279-1971 4663 control room - ana lysis FSAR 7.4.3.3.2.2 Alternate shutdown indication system Designed to function during and 4530 power generation - design basis after control room fire FSAR 7.4.3.3.2.3 Alternate shutdown indication system 10CFR50, App. A, GDC 19 4532 guides, criteria and standards FSAR 7.4.3.3.2.3 Alternate shutdown indication system IEEE 279-1971 4533 guides, criteria and standards FSAR 7.4.3.3.2.3 Alternate shutdown indication system IEEE 323-1974 4534 guides, criteria and standards FSAR 7.4.3.3.2.3 Alternate shutdown indication system IEEE 344-1975 4535
<
guides, critoria and standards FSAR 7.4.3.3.2.3 Al ternate shutdown indication system RG 1.22 4536 guides, criteria and standa rds FSAR 7.4.3.3.2.3 Alternate shutdown indication system BTP CMEB 9.5-1 4537 guides, criteria and standards FSAR 7.4.3.3.2 Alternate shutdown indication system BTP CMEB 9.5-1, Req. C.S.C 4664 design bases information FSAR Table 7.5 Post-accident monitoring RG 1.97, Rev. 2 4541 2-1 i nst rumenta t i on FSAR Table 7.5 Information systems va riable Summa ry of design, qua li fication 4548 2-2 categories and interface requi rements FSAR 7.5.2.3.1.3.A Description of info. system RG 1.75 4538 design criteria for category i FSAR 7.5.2.3.1.3.8 Description of info, system RG 1.32 4539 design criteria for category i FSAR 7.5.2.3.1.3.K Description of info, system RG 1.105 4540 design criteria for category l FSAR 7.5.2.3.2.3.J Description of info. system RG 1.105 4807 design criteria for category 2
.
_
_
_
-
. _.
.
TABLE 1.
COMMITMENTS (continued)
Numb 3r Sou rce Section Subject Document /Featura RIra rancs FSAR 7.5.4 Info. system important to sa fe ty NUREG-0737 4542 additional info.
{
FSAR 7.5.5.1 Bypassed and inoperable status RG 1.47 4543 indication for ESFS description FSAR 7.5.5.1 Bypassed and inoperable status IEEE 384 4544 indication for ESFS description FSAR 7.5.5.1 Bypassed and inoperable status RG 1.75 4545 indication for ESFS description FSAR 7.5.5.2 Bypassed and inoperable status RG 1.47 4546 indication for ESFS conf. to RG 1.47 i
FSAR 7.5.5.3 Bypassed and inoperable status BTP-ICSB-21, Rev. 2 4547 indication for ESFS conf. to BTP-ICSB-21 FSAR 7.6.1.2 I&C power supply system analysis IEEE 308-1974 4549
'
FSAR 7.6.1.2 I&C power supply system analysis RG 1.6 4808
!
FSAR 7.6.1.2 l&C power supply system analysis Vi ta l AC-no single fa i lure in the 4809
I&C power supply system or associated power supplies can cause loss of power to more than one of the redundant loads FSAR 7.6.2.2 Residual heat remova l IEEE 279-1971 4550 isolation valves analysis Sect. 4.10, 4.12, 4.15 FSAR 7.6.2.2 Residual heat removal IEEE 384-1974 4551 isolation valves analysis FSAR 7.6.4.A Interlock ci rcuits of MOV's in BTP ICSB4 5035 ECCS accumulator lines FSAR 7.6.6.7.A Interlocks isolating sa fety system RG 1.29 4552 from non-safety system analysis F SAR 7.6.6.7 A Interlocks isolating safety system RG 1.62 4553 f rom non-safety system ana lysis i
j FSAR 7.6.6.7.A interlocks isolating safety system RG l.75 4554 i
f rom non-safety system ana lysis
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TABLE 1 COMMITMENTS (centinued)
Number Sou rce Section Subjsct Document /Fcatu ra R;fa rcncs FSAR 7.6.6.7.A Interlocks isolating safety system IEEE 384 4555 f rom non-safety system analysis FSAR 7.6.6.7.B Interlocks isolating safety system IEEE 279-1971 4556 f rom non-safety system analysis FSAR 7.6.7.1 Interlocks for RCS press, control BTP RSB 5-2 4557 during low-temp. oper ana lysis of interlock FSAR 7.6.7.1 Interlocks for RCS press control IEEE 279-1971 4558 during low-temp. oper. ana lysis of interlock ISAR 7.7.2.1 Separation of protection and IEEE 279-1971, Sect. 4.7 4560 control system FSAR 7.7.2.2 Reactor shutdown with control rod s 10CFR50, App.
A, CDC 25 4561 FSAR 7.7.2.2 Reactor shutdown with control rod s 10CFR50, App. A, CDC 23 4562 FSAR 7.7.2.7 Co re coo l i ng moni to r NUREG 0737, item II.F.2 4564 FSAR 7.7.2.8 Reactor vessel level NUREG-0737, item II.F.2 4565 instrumentation system FSAR 8.3.1.1.8.B Design criteria class 1E equipment All class 1E motor-operated valve 4822 actuators are specified with accelerating capability at 75%
nameplate FSAR 8.3.1.4.3.C Cable routing Separation - within panels and
control boa rds, 6 in, minimum spatial sepa ration or ba rriers a re provided between components or cables of diff, sepa ration groups Note 1.
F SAR 7. 3. 3.1.2 Comm i tment ( Re f.
No. 4642) erroneously lists IEEE 279-1971, Section 4.12; shculd be Section 4.17.
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TABLE 2.
CDMMITMENT IMPLEMENTATION Document / Feature Number Design Last Reference Separation - within panels DC-1601, Rev. 1 82.00 and control boards, 6 in, minimum spatial separation or barriers provided between components or are cables of diff. separation groups (FSAR 8.3.1.4)
RG 1.29, Rev. 3 DC-1000-J, Rev. 4 126.00 (FSAR 1.9.29)
DC-1005, Rev. 1 RG 1.97, Rev. 2 (FSAR DC-1000-J, Rev. 4 178.00 1.9.97)
DC-1601, Rev. 1 0C-1623, Rev. 1 0C-2701, Rev. O RG 1.44 (FSAR 6.1.1.1.3)
X4AZ01, Div. P1 259.00
>
Rev. 14, P. 15 para. G ASME III, NC-2160 &
DC-1204, Rev. 2 322.00 NC-3120 (FSAR 6.1.1.1.1)
DC-1205, Rev. 2 DC-1206, Rev. 2 DC-2415, Rev. 1 X5AG15-4, Rev. 11 X4P0003-C.4, Rev. 2 10CFR100 (FSAR 1.2.6)
DC-1000-J, Rev. 2 673.00 ASME III (FSAR 3.10.B.3)
0C-1017, Rev. 4 1223.00 DC-1000-J, Rev. 4 DC-1005, Rev. 1 RG 1.53, Rev. 0 (FSAR DC-1000-J, Rev. 4 1551.00 1.9.53)
DC-1601, Rev. 1 DC-1605, Rev. 1 DC-1620, Rev. 2 IEEE 379-1972 (FSAR See Remarks 1552.00 1.9.53)
IEEE 279-1971, Sect. 4.2 DC-1009, Rev. 2 1553.00 (FSAR 1.9.53) RG-1.53 RG 1.84, Rev. 20 DC-1000-J, para.
1565.00 (FSAR 1.9.84)
18, Rev. 4 RG 1.141, Rev. O DC-2415, Rev. 1 1595.00 (FSAR 1.9.141)
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TABLE 2.
COMMITMENT IMPLEMENTATION (continued)
Number Document / Feature Number Design Last Reference RG 1.7 Control of DC-1513, Rev. 0 2067.00 combustible gas conc. in conta.
following loca (FSAR 6.2.5.1)
NUREG-0737, II.E.1.1, DC-1604, Rev. 1 2307.00 II.E.1.2 (FSAR 10.4.9)
DC-1601, Rev. 1 2316.00 DC-1620, Rev. 2 DC-1624, Rev. 2 RG 1.29 (FSAR 6.2.2.1)
DC-1005, Rev. 1 2398.00 DC-1010, Rev. 5 All power supplies and DC-2415, Rev. 1 2433.00 control fcts necessary for conta. isolation are class 1E (FSAR 6.2.4.5)
BTP ICSB-18 (FSAR DC-1009, Rev. 2 2817.00 6.3.2.2.16)
IEEE 279-1971 (FSAR 7.1)
DC-1620, Rev. 2 4359.00 DC-1624, Rev. 2 ESFAS must have
_
DC-1605, Rev. 1 4479.00 provisions-in control DC-1620, Rev. 2 room for manually initiating functions of ESF system (FSAR 7.1.2.1)
BTP.ICSB-21, Rev. 2 DC-1625, Rev. 0 4547.00 (FSAR 7.5.5.3)
DC-1201, Rev. 2 4557.00 IEEE 279-1971 (FSAR DC-1000-J, Rev. 4 4616.00 7.2.2.2)
DC-1605, Rev. 1 All class 1E motor Spec. X5AC01, 4822.00 operated valve actuators App. AD-5, Rev. 2 are specified with acces.
capability at 75% nameplate (FSAR 8.3.1.1)
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TABLE 2.
COMMITMENT IMPLEMENTATION (continued)
Document / Feature Construction Last Reference ASME Code, Sect. III, 1977 PPP IX-40, Rev.
5026.00 edition thru winter 1977 09/09/85, X-18, addenda Rev. 11/15/85 ANSI 831.1 code, 1977 PPP IX-40, Rev.
5027.00 edition thru winter 1977 09/09/85, X-18, addenda Rev. 11/15/85 AWS D1.1-1975 PPP.IX-40, Rev.
5028.00 09/09/85, X-18, Rev. 11/15/85 Note 1.
Commitment Ref. Nos. 1491, 2405, and 2406 do not apply to this module and should be disregarded.
Note 2.
Commitment Ref. No. 2398 lists DC-1010, Rev. 4 as the current version of the design document, it should be Rev. 5.
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TABLE 3.
ADDITIONAL AUDITS Audit No.
Audit Date Subject Findings VH-III-1 12-02-80 Drawings Note 1 VH-III-2 06-09-80 Supplier Dwgs Note 1 VH-III-3 03-04-80 Specifications Note 1
.VH-III-3 07-21-81 Specifications Note 1 VS-85-120 02-02-85 Miscellaneous Note 1 VS-85-123 03-29-85 Document Control Note 1 VS-85-124 03-29-85 Document Control Note 1 VS-85-125 03-29-85 Design Changes Note 1 VS-85-126 03-29-85 Design Review Note 1 VH-III-3 10-20-82 Specifications 82-15 VH-III-3 06-24-83 Specifications 83-24 VH-III-1 11-21-83 Drawings 83-39 VH-III-1 01-25-84 Drawings
~83-54 VH-III-3 03-29-84 Specifications 84-49 VH-III-16 08-15-84 Design Changes 84-56 VH-IV-1 02-25-85 Specifications 85-69 VH-III-16 02-25-85 Specifications 85-71 VH-III-8 03-07-85 Design Review 85-72 VH-III-17 08-22-85 Design Changes 85-84 VH-III-19 10-25-85 Miscellaneous 85-91 Note 1 - These audits were not selected for review; therefore, no information regarding these findings was examine '
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TABLE 4.
INSTRUMENTS SELECTED BY NRC REVIEWERS FOR WALKDOWN Instrument No.
System / Function FT-5150 Flow Transmitter for Auxiliary Feedwater (AFW) Flow from Turbine-Driven AFW Pump PI-0977 Pressure Transmitter for Safety Injection Pump Suction Pressure LT-0990 Level Transmitter for RWST Level PT-1853 Component Cooling Water Pump Discharge Pressure PT-0536 Main Steam Line Pressure PT-0936 Containment Pressure HV-5120 Auxiliary Feedwater Flow Control Valve Operator and E/P (Electrical Signal to Air Pressure) Converter PI-0601 Reactor Heat Removal Pump Suction Pressure PI-0974 Containment Spray Discharge Pressure PT-1636 Service Water Pump Discharge Pressure PT-0405 Reactor Coolant System-Reactor Heat Removal System Recirculating Line LT-0461 Pressurizer Level Transmitter
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ACRONYMS American National Standards Institute ANSI
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American Society of Mechanical Engineers ASME
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L BPC
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Bechtel Power Company Code of Federal Regulations CFR-
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DCN
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Document Change Notices DR
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Deviation Report Electrical Quality Control l
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Field Change Request FCR
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Final Safety Analysis Report Georgia Power Company
GPC
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l HOE
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BPC, Home Office Engineering Instrumentation and Controls I&C
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Interim Design Changes Notices IDCN
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Independent Design Review IDR
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Office of Inspection and Enforcement IE
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Institute of Electrical and Electronic Engineers, Inc.
IEEE
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j IFI Inspector Followup Items
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Institute of Nuclear Power Operations INP0
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LT Level Transmitter
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Megawatts electrical MWE
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MWT
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Megawatts thermal NRC
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Nuclear Regulatory Commission Office of Nuclear Reactor Regulation NRR
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Nuclear Steam Supply System NSSS
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Project Field Engineering PI Pressure Indicator PFE
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Pullman Power Products PPP
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Project Reference Manual PRM
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Preliminary Safety Analysis Report PSAR
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PT Pressure Transmitter
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Quality Assurance QA
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Quality Control QC
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Regulatory Guide RRT Readiness Review Team
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Supplier Deviation Disposition Request SDDR
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Safety Evaluation Report SER
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SIE Self-Initiated Evaluation
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SIS Safety Injection System
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Unresolved Item URI
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Vogtle Electric Generating Plant V-SAMV
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Vogtle Electric Generating Plant VEGP
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VNP Vogtle Nuclear Plant
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