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[g UNITED STATES g/;#
g NUCLEAR REGULATORY COMMISSION ti f '
C WASHINGTON,0. C. 20555 i
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0CT 111934 L
Docket Nos. 50-424/425 MEMORANDUM FOR: Thomas M. Novak, Assistant Director I
for Licensing Division of Licensing FROM:
William V. Johnston, Assistant Director i
Materials, Chemical & Environmental T.echnology Division of Engineering l
SUBJECT:
GEORGIA POWER COMPANY, V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 & 2, ORAFT SAFETY EVALUATION REPORT Plant Name: Vogtle Electric Generating Plant, Units 1 & 2 Suppliers: Westinghouse, Southern Services, Bechtel Licensing Stage: OL Docket Numbers: 50-424/425 Responsible Branch & Project Manager:
LB-4, M. Miller e s, Reviewers:
M. R. Hum, B. Brown, (INEL) and L. Frank
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Description of Task: Review of the Preservice Inspection Program Reouested Completion Date:
Draft SER Input - October 1,1984 SER Input - April 1, 1985 Review Status: Applicant's Response Required Section 5.2.4 - Open Issue Section 5.4.2.2 - Open Issue Section 6.6 - Open Issue In accordance with a request from the Project Manager, the Inservice Inspection Section, Materials Engineering Branch, Division of Engineering, has reviewed the information in the FSAR through Amendirent 9 (August 1984) and the Pre-service Inspection (PSI) Program, Revision 0, dated April 19, 1984. Our review of the PSI Program has determined that the document is not complete.
In addition, the Applicant has not committed to perform a preservice examina-tion of the full length of each tube in each steam generator based on the Stand'ard Technical Specification, NUREG-0452, Revision 2.
We have prepared the attached draft SER input to Sections 5.2.4, 5.4.2.2'and 6.6 that describes the conclusions from our review and the information that the Applicant should provide to resolve the open issues before the publication
Contact:
M. R. Hum i
X-28482
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m Thomas M. Novak OCT 1119s.;
of the SER.
Information on the steam generators that is related to GANE proposed Contention 11 is provided in draft SER Section 5.4.2.2.
Although the Applicant submitted a separate PSI Program for Units 1 and 2, the staff is focusing its review on Unit I considering the significant differences in the construction completion schedules. The staff is taking this approach because Applicants with multiple units have frequently updated the entire PSI Program for the second unit after completion of the initial unit.
The MTEB reviewers and our consultant are prepared to discuss the resolu-tion of these issues with the Applicant.
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f Wilitam V. Johnston; Assistant Director aterials, Chemical & Environmental Technology Division of Engineering
Attachment:
As Stated cc:
R. Vollmer r-D. Eisenhut f
W. Johnston T. Novak E. Sullivan S. Pawlicki M. Miller B. D. Liaw C. Y. Cheng W. Hazelton R. Klecker L. Frank j
J. Blake, Region II B. Brown, INEL G. Johnson M. Hum i
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GEORGIA POWER CO.
V0GTLE ELECTRIC GENERATING PLANT-UNITS 1 AND 2 DOCKET NUMBERS 50-424/425 2
DRAFT SAFETY EVALUATION REPORT t
P MATERIALS ENGINEERING BRANCH INSERVICE INSPECTION SECTION 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing This section was prepared with the technical assistance of 00E contractors from the Idaho National Engineering Laboratory.
5.2.4.1 Compliance with the Standard Review Plans
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The July 1981 Edition of the " Standard Review Plan for the Review of Safety A'nalysis Reports for Nuclear Power Plants'.' (NUREG-0800) includes Section 5.2.4,
" Reactor Coolant Pressure Boundary Inservice Inspection and Testing." The Vogtle Electric Generating Plant (VEGP) review is continuing because the Applicant has not completed the Preservice Inspection Program (PSI) and examinations.
The staff review to date was conducted in accordance with Standard Review Plan (SRP) Section 5.2.4 except as discussed below.
Paragraph II.4, " Acceptance Criteria, Inspection Intervals," has not been i
reviewed because this area applies only to inservice inspections (ISI), not to PSI.
This subject will be addressed duri'ng review of the ISI program after licensing.
1 Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed.
The Applicant committed in the FSAR to incorporate ASME Code Article IWB-3000, " Acceptance Standards for Flaw Indications," into the PSI Program.
However, ongoing NRC generic activities and research projects indicate f
that the presently specified ASME Code procedures may not always be capable of detecting the acceptable size flaws spect fled in the I'43-3000 acceptance 10/09/84 5-1 V0GTLE SER INPUT SEC 5.2.4 1*
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standards.
For example, ASME Code procedures specified for volumetric exam-ination of the reactor vessel, bolts and studs, and piping have not proven to be capable of detecting the acceptable size flaws in all cases.
The staff will continue to evaluate the development of new or improved procedures and will require that these improved procedures be made a part of the inservice examination requirements. The Applicant's repair procedures based on ASME Code Article IW8-4000, " Repair Procedures," have not been reviewed.
Repairs L
are not generally necessary in the PSI program.
This subject will be addressed during the staff review of the ISI program.
Paragraph II.7, " Acceptance Criteria, Code Exemptions," has not been completed because the Applicant has not listed in the FSAR or the PSI Program any Code exemptions as permitted by the criteria in IWB-1220.
The SRP requires that the Applicant list these exemptions, if used.
Paragraph II.8, " Acceptance Criteria, Relief Requests," has not been completed because the Applicant lias not identified all limitations to examination.
Specific areas where ASME Code examination requirements cannot be met will be iden'tified as performance o'f the PSI progresses.
The complete evaluation of the PSI program will be presented in a supplement to this Safety Evaluation Report (SER) after the Applicant submits the required examination information, identifies all plant-specific areas where ASME Code Section XI requirements cannot be met, and provides a supporting technical justification.
5.2.4.2 Examination Requirements General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary,"
Appendix A of 10 CFR Part 50 requires, in part, that components of the reactor coolant pressure boundary be designed to permit periodic examination and testing o'f important areas and features to assess their structural and leak-tight integrity.
To ensure that no deleterious defects develop during service, selected welds and weld heat-affected-zones (HAZ) will be examined periodically j
at Vogtle Electric Generating Plant.
10/09/84 5-2 VCGTLE SER INPUT SEC 5.2.4 I
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The design of the ASME Code Class 1 and 2 components of the reactor coolant pressure boundary incorporates provisions for access for inservice examinations, as required by Subarticle IWA-1500 of Section XI of the ASME Code. ' Paragraph 50.55a(g), 10 CFR Part 50, defines the detailed requirements j
for the preservice and inservice programs for light water-cooled nuclear power
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facility components.
Based upon the construction permit date of June 28, j
1974, components (including supports) which are classified as ASME Code Class 1 f
and 2 shall meet the preservice examination requirements set forth in editions of Section XI of the ASME Code and addenda in effect 6 months prior to the date of issuance of the construction permit.
The components (including supports) may meet the requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of 10 CFR 50.55a,
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subject to the limitations and modifications listed therein.
The initial ISI program must comply with the requirements of the latest edition and addenda of Section XI of the ASME Code in effect twelve months prior to the date of
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issuance,of the operating' license, subject to the limitations and modifi-cations listed in Paragraph 50.55a(b) of 10 CFR Part 50.
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i 5.2.4.3 Evaluation of Compliance with 10 CFR 50.55a(g)'
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Review has been completed on the information presented in the FSAR through Amendment 9 dated August 1984 and the Preservice Inspection Program, Revision 0, i
dated April 19, 1984.
The Applicant states that based on the construction e
f permit date for Vogtle, the Preservice Inspection Program is required to meet f
ASME Code Section XI, 1971 Edition through the Winter 1972 Addenda.
The Applicant has voluntarily updated the Preservice Inspection Program based on k,
ASME Code Section XI, 1980 Edition with Addenda through Winter 1980.
The use f' >
of later referenced Code editions is acceptable as specified by 10 CFR 50.55a(g).
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The PSI Program for the reactor coolant pressure boundary systems and components 1
has been reviewed. As the Ap.nitcant stated in the FSAR, these systems and f}
components are included for examinations per the applicable Code requirements.
l The staff established technical positions in the FSAR Questions, some of which I
are resolved in the PSI Program.
The following items require further input or clarification from the Applicant:
10/09/84 5-3 V0GTLE SER IflPUT SEC 5.2.4 l-
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A.
The PSI Program should contain a list of components subject to examination and a description of the welds exempted from examination in accordance with IWB-1220 and IWC-1220 and include the criteria for exemption.
In j
addition, the examination isometric drawings are necessary for the staff to determine the acceptability of the sample of welds required to be examined.
l 8.
In respone to FSAR Questions 250.2 and 250.3, the Applicant states that t
ultrasonic examination procedures have been or are being developed for the examination of (1) the reactor coolant pipe and fittings fabricated from SA351, Grade CF8A (centrifugal cast stainless steel) and (2) the j_
reactor vessel which address the degree of compliance with Regulatory Guide 1.150 and discuss the near-surface examinations, the resolution j
~with regards to detection of actual flaws, and the use of electronic j
gating as related to the volume of material near the surface of the F
reactor pressure vessel (RPV) that may not be examined.
The staff requests that the Applicant provide these examination procedures for the
' staff to complete this review.
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The PSI examination of the steam generator tubes should be performed in accordance with N REG-0542, Revision 2, and Regulatory Guide 1.83, j
Revision 1, as discussed in SER Section 5.4.2.2.
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1 The specific areas where the Code requirements cannot be met will be identified j
after the examinations are performed.
The Applicant has committed to identify l
all plant-specific areas where the Code requirements cannot be met and provide l
a supporting technical justification for requesting relief.
The SER input l
will be completed after the Applicant:
(1) Dockets an acceptable resolution to the above issues.
'(2) Dockets a complete and acceptable PSI Program.
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I (3) Submits all relief requests with a supporting technical justification.
The staff considers the review of the Vogtle Electric Generating Plant PSI Program an open issue subject to the Applicant providing the above items.
Evaluation of the response will be reported in the SER.
The initial Inservice Inspection Program has not been submitted by the Applicant.
This program will be evaluated after the applicable ASME Code
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edition and addenda can be determined based on Paragraph 50.55a(b) of 10 CFR
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Part 50, but before inservice inspection commences during the first refueling outage.
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5.2.4.5 Conclusions The conduct of periodic examinations and hydrostatic testing of pressure-retaining components of the reactor coolant pressure boundary, in accordance L
with.the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and 10 CFR Part 50, will provide
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h reasonable assurance that structural degradation or loss of leak-tight integ-rity occurring during service will be detected in time to permit corrective action before the safety functions of a component are compromised.
Compliance with the preservice and inservice examinations required by the Code and 10 CFR 3
Part 50 constitutes an acceptable basis for satisfying the inspection require-l ments of General Design Criterion 32.
5.2.4.6 References 1.
NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspection and Testing," Rev. 1, July 1981.
2.
Code of Federal Regulations, Volume 10, Part 50.
1 3.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1, 1980 Edition, through Winter 1980 Addenda.
10/09/84 5-5 VCGTLE SER IflPUT SEC 5.2.4 p
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Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations."
5.
NUREG-0452, Revision 2, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors."
6.
Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes."
5.4.2.2 Steam Generator Tube Inservice Inspection This section was prepared with the technical assistance of 00E contractors from the Idaho National Engineering Laboratory.
l 5.4.2.2.1 Compliance with the Standard Review Plans The July 1981 Edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," (NUREG-0800) includes Section 5.4.2.2 g
" Steam Generator Tube Inservice Inspection." The Vogtle FSAR through Amendment 9 dated August 1984 and the Preservice Inspection Program, Revision 0, dated April 19, 1984, was reviewed in accordance with this section of the Standard Review Plan (SRP).
The results of this review are summarized below.
The SRP Acceptance Criteria recommend that the Applicant perform examinations based on Regulatory Guide 1.83 and the applicable Standard Technical Specifica-tions.
The Applicant has not committed to perform the examination of the steam generator tubes in accordance with the recommendations of both Regulatory Guide 1.83 and the technical requirements of NUREG-0452, Revision 2.
5.4.2.2.2 Evaluation of the Inspection Program General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary,"
Appendix A of 10 CFR Part 50 requires, in part, that components of the reactor coolant pressure boundary be designed to permit periodic examination and testing of important areas and features to assess their structural and leak-tight integrity. The steam generators at VEGS have been designed to meet the ASME s
P 10/09/84 5-6 V0GTLE SER INPUT SEC 5.2.4
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Boiler and Pressure Vessel Code requirements for Class 1 and 2 components.
1 Provisions also have been made to permit inservice inspection of the Class 1 and 2 components, including individual steam generator tubes. The design aspects that provide access for examination and the proposed inspection program must comply with the requirements of Section XI of the ASME Code, follow the recommendations of Regulatory Guide 1.83, Revision 1, " Inservice Inspection of i
Pressurized Water Reactor Steam Generator Tubes," and NUREG-0452, Revision 2, "Standa'rd Technical Specifications for Westinghouse Pressurized Water Reactors,"
with respect to the examination methods to be used, provisions for a baseline examination, selection and sampling of tubes, inspection intervals, and actions to be taken in the event that defects are identified.
The Applicant has not committed to perform a preservice' examination of the full length of each tube in each steam generator using eddy-current techniques to establish the baseline condition of the tubing based en the Standard Tech-nical Specification, NUREG-0452, Revision 2.
The examination is to be performed j
prior to initial power operation using the equipment and techniques expected to
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be used during subsequent inservice examinations.
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The Vogtle steam generators are the latest Westinghouse models, designated Model F, which incorporate multiple features to minimize operating problems i
associated with crevice corrosion, stress corrosion cracking, flow-induced l
vibration, and denting.
These features include:
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1.
Type 405 ferritic stainless steel quatrefoil tube support plate to eliminate denting, i
2.
Thermally treated Inconel 600 tubing and stress relief of the inner-most i
eight rows of the tube bundle to reduce the potential for stress corrosion
- cracking, i
3.
Expansion of the tubes to the full depth of the tubesheet to eliminate l
crevices and potential for crevice corrosion,
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10/09/84 5-7 V0GTLE SER INPUT SEC 5.2.4 1
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A flow baffle plate above the tubesheet to direct lateral flow across the tubesheet surface and thus minimize the number of tubes exposed to sludge
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and potential for corrosion attack, and 5.
An improved blowdown system to increase blowdown capacity to minimize sludge build-up, t
i Vibration induced wear which has been experienced in the preheater section of Model D's and E Westinghouse steam generators, prior to modifications, will
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not be a problem in Vogtle Model F steam generators since this model does not have a preheater section.
l The staff considers the examination of the steam generators an open issue, subject to the applicant committing to the applicable provisions of the Standard Technical Specification.
The staff will complete the review and report the results in a supplement to the SER after the actual Technical Specifications are established.
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5.4.2.2.3 Conclusions Conformance with Regulatory Guide 1.83, NUREG-0452 and the inspection requirements of Section XI of the ASME Code consitutes an acceptable basis for meeting, in part, the requirements of General Design Criterion 32.
5.4.2.2.4 References 1.
NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary i
Inservice Inspection and Testing," Section 5.4.2.2, " Steam Generator Tube L
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Inservice Inspection," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components," July.1981.
i 2.
Code of Federal Regelations, Volume 10, Part 50.
I 3.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, 1980 Edition through Winter 1980 Addenda, e
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4.
NUREG-0452, Revision 2, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors."
j; 5.
Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes."
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10/09/84 5-9 VCGTLE SER INPUT SEC 5.2.4 1
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=--,mn-E 6.6 Inservice Inspection of Class 2 and 3 Components r
This section was prepared with the technical assistance of DOE contractors from the Idaho National Engineering Laboratory.
6.6.1 Compliance with the Standard Review Plans The July 1981 Edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (NUREG-0800) includes Section 6.6, l
" Inservice Inspection of Class 2 and 3 Components." The Vogtle review is I
continuing because the Applicant has not completed the PSI program and examina-tions.
The staff review to date was conducted in accordance with Standard j
' Review Plan Section 6.6 except as discussed below.
4 Paragraph II.4, " Acceptance Criteria, Inspection Intervals," has.not been
[h reviewed because this area appifes only to inservice inspection (ISI) not to PSI.
This subject will be addressed during review of the ISI program after Ifcensing.
Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed.
The Applicant committed in the FSAR to incorporate ASME Code e
f Article IWC-3000 " Acceptance Standards for Flaw Indications," into the PSI program.
However, ongoing NRC generic activities and research projects indicate
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that the presently specified ASME Code procedures may not always be capable of detecting the acceptable size flaws specified in these standards.
For example, j
ASME Code procedures specified for volumetric examination of vessels, bolts and studs, and piping have not proven to be capable of detecting acceptable
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size flaws in all cases.
The staff will continue to evaluate the development of new or improved procedures and will require that these improved procedures be made a part of the inservice examination requirements.
The Applicant's repai.r procedures based on ASME Code Articles IWC-4000 and IWD-4000, " Repair t
Procedures," have not been reviewed.
Repairs are not generally necessary in the PSI program.
This subject will be addressed during review of the ISI program.
10/09/84 6-1 V0GTLE SER INPUT SEC 6.6 1
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Paragraph II.8, " Acceptance Criteria, Code Exemptions," has not been completed
.s because the Applicant has not listed in the FSAR or the PSI Program any Code exemptions as permitted by the criteria in IWC-1220.
The SRP requires that the Applicant list these exemptions, if used.
Paragraph II.9, " Acceptance Criteria, Relief Requests," has not been completed i
because the Applicant has not identified the limitations to examination.
b Specific areas where ASME Code examination requirements cannot be met will be identified as the PSI progresses.
The complete evaluation of the PSI program will be presented in a supplement to the Safety Evaluation Report (SER) after the Applicant submits the required examination information and identifies all plant-specific areas where ASME Code Section XI requirements cannot be met and provides a supporting technical justification.
i 6.6.2 Examination Requirements General Design Criteria 36, 39, 42, and 45, Appendix A of 10 CFR Part 50,
-3, require, in part, that the Class 2 and 3 components be designed to permit appropriate periodic inspection of important components to ensure system l
integrity and capability.
Paragraph 50.55a(g) of 10 CFR Part 50 defines the detailed requirements for the preservice and inservice inspection programs for
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light water-cooled nuclear power facility components.
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l Based upon the construction permit date of June 28, 1974, this paragraph of j
the regulati.ons requires that a preservice inspection program for Class 2 components (including supports) shall meet the preservice examination require-
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ments set forth in editions of Section XI of the ASME Code and addenda in effect 6 months prior to the date of issuance of the construction permit.
The components (including supports) may meet the requirements set forth in subse-quent editions of this Code and addenda which are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein.
The initial ISI program must comply with the requirements of the latest edition i
and addenda of Section XI of the ASME Code in effect twelve months prior to the date of issuance of the operating license, subject to the limitations and I
modifications listed in Paragraph 50.55a(b) of 10 CFR Part 50.
L 10/09/84 6-2 V0GTLE SER IrlPUT SEC 6.6
p 6.6.3 Evaluation of Compliance with 10 CFR 50.55a(g)
Review has been completed on the information presented in the FSAR through Amendment 9 dated August 1984 and the Preservice Inspection Program, Revision 0, dated April 19, 1984. The Applicant stated that based on the construction permit date for Vogtle, the Preservice Inspection Program is required to meet ASME Code Section XI, 1971 Edition through Winter 1972 Addenda. The Applicant has voluntarily updated the Preservice Inspection Program based on ASME Code Section XI, 1980 Edition with Addenda through Winter 1980.
The use of later referenced Code editions is acceptable as specified by 10 CFR 50.55a(g).
The PSI Program for the Class 2 and 3 components has been reviewed.
As the Applicant stated in the FSAR, these systems and components are included for examinations per the, applicable Code requirements.
The staff established technical positions in the FSAR Questions, some of which are resolved in the PSI Program. The following items require further input or clarification from the Applicant:
h A.
The PSI Program should contain a list of components subject to examination
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and a description of the welds exempted from examination in accordance with IWC-1220 and include the criteria for exemption.
In addition, the examinatico isometric drawings for ASME Class 2 components are necessary fer the Kpf to determine the acceptability of the sample of welds required to be examined.
B.
10 CFR 50.55a(b)(2)(iv) requires that ASME Code Class 2 piping welds in the Residual Heat Removal (RHR) Systems, Emergency Core Cooling (ECCS)
Systems, and Containment Heat Removal (CHR) Systems shall be examined.
These systems should not be completely exempted from'preservice volu-metric examination based on Section XI exclusion' criteria contained in IWC-1220.
To satisfy the inspection requirements of General Design Crteria 36, 39, 42, and 45, the preservice inspection program must include volumetric examination of a representative sample of welds in the RHR, ECCS and CHR Systems.
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The specific areas where the Code requirements cannot be met will be identified after the examinations are performed.
The Applicant has committed to identify all plant-specific areas where the Code requirements cannot be met and provide I
a supporting technical justification for requesting relief.
The SER input will be completed after the Applicant:
(1) Dockets an acceptable resolution to the above issues.
(2) Dockets a complete and acceptable PSI Program.
t (3) Submits all relief request with a supporting technical justification.
The staff considers the review of the Vogtle Electric Generating Plant PSI Program an open issue subject to the Applcant providing the above items.
Evaluation of the response will be reported in the SER.
The initial Inservice Inspection Program has not been submitted by the Applicant.
This program will be evaluated after the applicable ASME Code Edition and Addenda can be determined based on Paragraph 50.55a(b) of 10 CFR c
I Part 50, but before inservice inspection commences during the first refueling outage.,
6.6.5 Conclusions Compliance with the preservice and inservice inspections required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and
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10 CFR Part 50 constitutes an acceptable basis for satisfying applicable l
requirements of General Design Criteria 36, 39, 42, and 45.
6.6.6 References 1.
NUREG-0800, Standard Review Plan, Section 6.6, " Inservice Inspection of Class 2 and 3 Components," Rev.1, July 1981.
l 2.
Code of Federal Regulations, Volume 10, Part 50.
3.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1,1980 Edition through Winter 1980 Addenda.
10/09/84 6-4 VCGTLE SER INPUT SEC 6.6 h
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