ML20209C638
| ML20209C638 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 11/25/1983 |
| From: | Johnston W Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1474 NUDOCS 8312070279 | |
| Download: ML20209C638 (23) | |
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!;C'l 2 5 1983 Docket No. 50-508 MEMORANDUM FOR: Thomas M. Novak, Assistant Director l for Licensing Division of Licensing FROM: William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering i
SUBJECT:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS),
WNP-3, DRAFT SAFETY EVALUATION REPORT - PRESERVICE/
~ INSERVICE . INSPECTION PROGRAMS Plant Name: Washington Nuclear Power, Unit 3 '
Suppliers: Combustion Ergineering and EBASCO Docket No: 50-508 Responsible Branch & Project Manager: LB #3, A. Vietti Reviewer: M. R. Hum (B. Brown, INEL)
Description of Task: Draft Safety Evaluation Report on PSI /ISI Programs Requested Completion Date: October 28, 1983 Review Status: Open Item In accordance with your request, the Inservice Inspection Section of the Materials Engineering Branch, Division of Engineering has prepared the attached Draft SER for Sections 5.2.4, 5.4.2.2, and 6.6. The Applicant has not responded to our acceptance review questions and has not submitted the Preservice Inspection Program for review.
Since the Applicant has not provided substantive informaticn regarding the preservice examinations since announcing a suspension to construction, we consider the review of preservice inspections program an open issue.
Based on the review of the available infonnation in the FSAR, we have prepared the technical positions as Attachment 2 to request additional information to complete our review.
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William V. Johnston, Assistant Director Materials, themical & Environmental Technology Division of Engineering Attachneents: As stated cc: See Page 2 /
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ATTACHMENT I l
WASHINGTON PUBLIC POWER SUPPLY SYSTEM WASHINGTON NUCLEAR PROJECT NO. 3 j DOCKET NUMBER 50-500 DRAFT, SAFETY EVALUATION REPORT .
PRESERVICE/ INSERVICE INSPECT _ ION PROGRAM 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 DOE contractorsfromtheIdahoNationalEngina$ingLaboratory.
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 Analysis Reports for Nuclear Power Plants," (NUREG-0800) includes Section 5.2.,4, " Reactor Coolant Pressure Boundary Inservice Inspection and Testing." The Washington Nucleac Project No. 3 (WNP-3) review is continuing because the Applicant has not submitted a Preservice Inspection (PSI) Program and has not completed the PSI examinations. In FSAR Table 1.8-3, the Applicant has committed to comply with the Standard Review Plan (SRP) 5.2.4 acceptance criteria. The staff review to date was conducted in accordance I
with SRP Sectio'n 5.2.4 except as discussed,below. ..
l Paragraph II.3, " Acceptance Criteria, Examination Categories and Methods," will be' reviewed when the complete PSI Piogram has been received'from the Applicant.
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. s p g Paragraph II.4, " Acceptance Criteria, Inspection Intervals," has not been reviewed because this area applies only to inservice inspections (ISI), not to PSI. This subject will be addressed during review of the ISI program after licensing.
Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed. The Applicant committed in the FSAR to incorporate ASME Code Section XI, Article IWB-3000, " Standards for Examination Evaluation," into the PSI Program. However, ongoing NRC generic activities and research projects indicate that the presently specified ASME Code procedures may not always be capable of detecting the acceptable size flaws specified in the IWB-3000 acceptance standards. For example, ASME Code procedures specified for volumetric examination of reactor vessels, 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 develop-ment 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 Section XI, Article IWB-4000, " Repair Procedures," have not been reviewed. Repairs 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," will be reviewed when the completed PSI Program Plan is submitted by the ,
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Paragraph II.8, ' Acceptance Criteria, Relief Requests," has not been completed because the Applicant has not identified all limita-tions to examination. Specific areas where ASME Code examination requirements cannot be met will be identified as performance of 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 l
General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary," Appendix A of 10 CFR Part 50 requires, in i
part, that components which are part of the reactor coolant pres-
! sure boundary be designed to permit periodic examination and testing of important areas and features _to assess their structural and leak-tight integrity. To ensure that no deleterious detects develop during service, selected welds and weld heat-affected-zones (HAZ) will be examined periodically. The design of the ASME Code Class 1 l
and 2 components of the reactor coolant pressure boundary incor-porates provisions for access for inservice examinations, as required by Paragraph IWA-1500 of Section XI of the ASME Code.
Section 50.55a(g), 10 CFR Part 50, defines the detailed require-ments for the preservice and inservice programs.
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Based upon the construction permit date of April 11, 1978, this section of the regulations requires that a preservice inspection program be developed and implemented using at least the Edition and Addenda of Section XI of the ASME Code applied to the con-struction of the particular components. The components (including supports) may meet requirements set forth in subsequent 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.
TheinitialISIprogrammustcomplywiththerequirementsgf the latest Edition 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 modifications listed in Section 50.55a(b) of 10 CFR Part 50.
5.2.4.3 Evaluation of Compliance with 10 CFR 50.55a(g) for Washington Nuclear Project No. 3 Review has been completed on the information presented in the FSAR through Amendment 3 dated April 1983. The preservice examination on the piping and components, except NSSS components, will be examined in accordance with the requirements of the 1977 Edition of ASME Code Section XI with Addenda through Summer 1978.
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The NSSS components will be examined in accordance with the requirements of the 1974 Edition of ASME Code Section XI with Addenda through Summer 1975 except that the steam generator tubing will be examined in accordance with ASME Code Section XI 1980 Edition with Addenda through Winter 1980.
The Preservice Inspection (PSI) Program for systems and compo-nents within the reactor coolant pressure boundary has not been received. However, the Applicant has stated in the FSAR that these systems and components will be examined per the applicable Code requirements. Based on the review of the FSAR, the staff has established technical positions that should be included in the PSI Program.
The Applicant has committed to identify all plant-specific areas where the Code requirements cannot be met after the examinations are performed and provide a supporting technical justification Yor requesting relief. The SER input will be completed after the Applicant:
l (1) Dockets a complete and acceptable PSI Program, l
(2) Submits the requested additional information regarding the l PSI /ISI program, and l .
l (3) Submits all relief requests with a supporting technical justification.
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The staff considers the review of the PSI Program an open issue subject to the Applicant providing an acceptable response to the above requirements. ..
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 Section 50.55a(b) of 10 CFR Part 50, but before inservice inspection commences during the first refueling outage.
5.2.4.4 Conclusions The conduct of periodic examinations and hydrostatic testing of pressure-retaining components of the reactor coolant pressure boundary, in accordance with the requirements of Section XI of the l ASME Code and 10 CFR Part 50, will provide reasonable assurance that structural degradation or loss of leak-tight integrity 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 Part 50 constitutes an acceptable basis for satisfying the inspection requirements of l
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5.2.4.5 References
- 1. NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspection and Testing," July 1981.
- 2. Code of Federal Regulations, Volume 10, Part 50.
- 3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1.
1974. Edition, through Summer 1975 Addenda 1977 Edition, through Summer,1978 Addenda
'1980 Edition, through Winter 1980 Addenda 5.4.2.2 Steam Generator Tube Inservice Inspection
! This section was prepared with the technical assistance of DOE contractors from the Idaho National Engineering Laboratory.
5.4.2.2.1 Compliance with the Standard Review Plans The July 1981 edition of the " Standard Review Plan for the Review l
of Safety Analysis Reports for Nuclear Power Plants" (NUREG-0800) includes Section 5.4.2.2, " Steam Generator Tube Inservice Inspection."
The FSAR was reviewed in accordance with this section of the Standard l
Review Plan (SRP). The results of this review are summarized below.
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The SRP Acceptance Criteria recommend that the Applicant perform
, examinations based on Regulatory Guide 1.83 and the applicable Standard Technical Specifications. The FSAR Table 1.8-3 states that compliance with Section b.4.2.2 of NUREG-0800 is under review and a compliance statement will be provided in a subsequent amendment.
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 wh.ich are part 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 have been designed to meet the ASME Boiler and Pressure Vessel Code requirements for Class 1 and 2 components.
Provisions also have been made to permit inservice inspection of the Class 1 and 2 components, including individual steam generator i
! tubes. The design aspects that provide access for examination and the proposed inspection program must comply with the requirements l
l of Section XI of the ASME Code with respect to the examination methods to be used, provisions for a baseline examination, i selection and sampling of tubes, inspection intervals, and actions to be taken in the event that defects are identified.
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The proposed inspection program must also follow the recommenda-tions of Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," and NUREG-0212, Revision 2, " Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors."
The Applicant, in Chapter 16 of the FSAR, has committed to develop the Technical Specifications using the guidance in the latest revision of NUREG-0212 as well as the approved CESSAR-F Technical Specifications. Based on the above, the staff considers the preservice examination of the steam generators an open issue subject to the Appli. cant docketing an inspection program that complies with the latest revision of NUREG-0212.
5.4.2.2.3 Conclusions Conformance with Regulatory Guide 1.83, NUREG-0212, and the inspection requirements of Section XI of the ASME Code constitutes an acceptable basis for meeting, in part, the requirements of General Design Criterion 32.
, 3 q 5.4.2.2.4 R'eferences
- 1. NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspection and Testing," Section 5.4.2.2, " Steam Generator Tube Inservice Inspection," July 1981.
- 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.
- 4. NUREG-0212, Revision 2, " Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors."
- 5. Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Steam Generator Tubes."
l 6.6 Inservice Inspection of Class 2 and 3 Components This section was prepared with the technical assistance of DOE contractors from the Idaho National Engineering Laboratory.
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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, " Inservice Inspection of Class 2 and 3 Components." The review is continuing because the Applicant has not submitted a Preservice Inspection (PSI) Program and has not completed the PSI examinations. The FSAR Table 1.8-3 states that compliance with Section 6.6 of NUREG-0800 is under review and a compliance statement will be provided in a subsequent amendment.
The staff review to date was conducted in accordance with Standard
, Review Plan Section 6.6 except as discussed below.
Paragraph II.3, " Acceptance Criteria, Examination Categories and Methods," will be reviewed when the completed PSI Program has ,
been received.
Paragraph II.4, " Acceptance Criteria, Inspection Intervals," has not been reviewed because this area applies only to inservice inspection (ISI) not to PSI. This subject will be addressed during review of the ISI program after licensing.
Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed. The Applicant committed in the FSAR to incorporate ASME Code Section XI, Articles IWC-3000 and IWD-3000,
" Standards for Examination Evaluation," into the PSI program.
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However, ongoing NRC generic activities and research projects indicate that the presently specified ASME Code procedures may not always be capable of detecting the acceptable size flaws specified in these standards. For example, ASME Code procedures specified for volumetric examination of vessels, bolts and studs, and piping have not proven to be capable of detecting 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 Section XI, Articles IWC-4000 and IWD-4000, " Repair 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.
Paragraph II.7, " Acceptance Criteria, Augmented ISI to Protect Against Postulated Piping Failures," has not been completed because this subject has not yet been addressed in the Applicant's PSI program. The Applicant's augmented ISI program will be reviewed after it is submitted.
Paragraph II.8, " Acceptance Criteria, Code Exemptions," will be reviewed for compliance to IWC-1220 when the Applicant's PSI Program has been received. Paragraph II.9, " Acceptance Criteria,
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Relief Requests," has not been completed because the Applicant has not identified the limitations to examination. 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 informa-tion and identifies all plant-specific areas where ASME Code Section XI requirements cannot be met and provides a supporting technical justification.
6.6.2 Examination Requirements General Design Criteria 36, 39, 42, and 45, Appendix A of 10 CFR Part 50 requires, in part, that the Class 2 and 3 components be designed to permit appropriate periodic inspection of important components to ensure system integrity and capability. Section 50.55a(g) of 10 CFR Part 50 defines the detailed requirements for the preservice and inservice inspection programs, til.
Based upon the construction permit date of April 11, 1978, this section of the regulations requires that a preservice inspection program for Class 2 and 3 components be developed and implemented using at least the Edition and Addenda of Section XI of the ASME Code applied to the construction of the particular components.
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The components (including supports) may meet the requirements set forth in subsequent 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 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 modifications listed in Section 50.55a(b) of 10 CFR Part 50.
6.6.3 Evaluation of Compliance with 10 CFR 50.55a(g) for Washington Nuclear Project No. 3 Review has been completed on the information presented in the FSAR through Amendment 3 dated April 1983. The Class 2 and Class 3 piping and components will receive preservice examinations in accordance with the requirements of the 1977 Edition of ASME i
Code Section XI with Addenda through Summer 1978. The secondary side of the steam generators will be examined in accordance with I
the requirements of the 1974 Edition of ASME Code Section XI with Addenda through Summer 1975.
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The Preservice Inspection (PSI) Program for the Class 2 and 3 components 11as not been received. However, the Applicant has stated in the FSAR that these components will be examined per the applicable Code requirements. Based on the review of the FSAR, the staff has established technical positions that should be included in the PSI Program. The Applicant has committed to identify all plant-specific areas where the Code require-ments cannot be met after the examinations are performed and provide a supporting technical justification for request'ing relief. The SER will be completed after the Applicant:
(1) Dockets a complete and acceptable PSI Program, (2) Submits the requested additional information regarding the PSI /ISI program, and (3) Submits all relief requests with a supporting technical justification.
The staff considers the review of the PSI Program an open issue subject to the Applicant providing an acceptable response to the above requirements.
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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 Section 50.55a(b) of 10 CFR Part 50, but before inservice inspection commences during the first refueling outage.
6.6.4 Conclusions Compliance with the preservice and inservice inspections required by the ASME Code and 10 CFR Part 50 constitutes an acceptable basis for satisfying applicable requirements of General Design Criteria 36, 39, 42, and 45.
6.6.5 References
- 1. NUREG-0800, Standard Review Plan, Section 6.6, " Inservice l
l Inspection of Class 2 and 3 Components", July 1981.
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- 2. Code of Federal Regulations, Volume 10, Part 50.
- 3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1.
1974 Edition through Summer 1975 Addenda 1977 Edition through Summer 1978 Addenda l
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ATTACHMENT 2 WASHINGTON PUBLIC POWER SUPPLY SYSTEM WASHINGTON NUCLEAR PROJECT NO. 3 DOCKET NUMBER 50-508 MATERIALS ENGINEERING BRANCH INSERVICE INSPECTION SECTION Review of the FSAR and Technical Positions Regarding the Preservice(PSI)/ Inservice (ISI) Inspection Programs 250.1 For completion of SER Sections 5.2.4 and 6.6, the staff requires that the PSI Program Plan be submitted for review. The PSI Program should include reference to the ASME Code Section XI Edition and Addenda that will be used for the selection of components for examinations, lists of the components subject to examination, a description of the components exempt from examination by the applicable Code, and the examination isometric drawings.
Paragraph 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 volumetric examination based on Section XI exclusion criteria contained in IWC-1220. To satisfy the inspection require-ments cf General Design Criteria 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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250.2 Plans for preservice examination of the reactor pressure vessel -
welds should address the degree of compliance with Regulatory Guide 1.150. In FSAR Section 1.8, Table 1.8-1, the Applicant indicates exceptions to Regulatory Guide 1.150. List the exceptions being taken and discuss the degree of compliance and the qualification of procedures to be used to assure finding service-induced flaws on the inside surface.
250.3 Describe the' measures taken to ensure that austenitic stainless steel piping welds are examined using effective techniques and the methods of assuring adequate examination sensitivity over the required examination volume. Discuss the preservice examina-tion criterie used to record, report, and plot geometric or metallurgical ultrasonic indications in the piping systems to assure correlation of baseline data with inservice inspection results.
The ASME Code,Section XI, 1977 Edition with Addenda through Summer 1978 and 1980 Edition specifies the use of Appendix III of Section XI for ferritic piping welds. If this requirement is not applicable (for example, for austenitic piping welds),
ultrasonic examination is required by Section XI to be conducted in accordance with the applicable requirements of Article 5 of b_ _
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.. l 3-Section V, as amended by IWA-2232. A technical justification is required if any alternatives are used. If Section XI, Appendix III, J
Supplement 7, will be used for the examination of austenitic piping welds, discuss the following:
. 1. All modifications permitted by Supplement 7
- 2. Methods of qualifying the procedure for examination through the weld (if complete examination is to be considered for examination conducted with only one side access).
When using either Article 5 of Section y or Appendix III of Section XI for examination of either ferritic or austenitic piping welds, the following should be incorporated:
- 1. Any crack-like indication, regardless of ultrasonic amplitude, discovered during examination of piping welds or adjacent base metal materials should be recorded and investigated by a Level II or Level III examiner to the extent necessary to determine the shape, identity, and location of the reflector.
- 2. The Owner should evaluate and take corrective action for the disposition of any indication investigated and found to be other than geometrical or metallurgical in nature.
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4 4-250.4 All preservice examination requirements defined in Section XI of the ASME Code that have been determined to be impractical must be identified and a supporting technical justification for requests for relief must be provided. The relief request submittal should include at least the following information:
- 1. For ASME Code Class 1 and 2 components, provide a table similar to IWB-2500 and IWC-2500 confirming that either the Section XI preservice examination was performed on the component or reifef is requested.
- 2. Where relief is requested for pressure retaining welds in the reactor vessel, identify the specific welds that did not receive a 100% preservice ultrasonic examination, and indicate the extent of the examination that was performed.
- 3. Where relief is requested for piping system welds (Examination Category B-J, C-F, and C-G), provide a list of the specific welds that did not received a complete Section XI preservice examination including drawing or isometric identification l
number, system, weld number, and physical configuration L
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'. r) 5-(e.g., pipe-to-nozzle weld, etc.). Indicate the extent of the preservice examination that was performed. When the volumetric examination was performed from one side of the weld, discuss whether the entire weld volume and the heat affected zone (HAZ) and base metal on the far side of the weld were examined. State the primary reason that a specific examination is impractical (e.g. , . support of component restricts access, fitting prevents adequate ultrasonic coupling on one side, component-to-component welds prevent ultrasonic examination, etc.). Indicate any alternative or supplemental examinations performed and methods of fabrication examination.
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