ML20052A911
| ML20052A911 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 04/06/1982 |
| From: | Hernan R Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8204290378 | |
| Download: ML20052A911 (24) | |
Text
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y APR 6 198?e Docket Hos. 50-329/330 APPLICAllT: Consumers Power Company FACILITY:
Hidland Plant, Units 1 and 2
SUBJECT:
SUlMARY OF fiEETIllG HELD WITil CONSUf1ERS POWER Off PRESERVICE IftSPECTI0ff PROGRA!I, fiARCH 15, 1982 On 11 arch 15,1982, f he flRC Staff and its consultant, Battelle florthwest, met in Bethesda, llaryland with Consumers Power Ccapany, and their Consultants, Southwest Research Inc. and General Physics Corporation, to discuss the Preservice Inspection Progran for the Ilidland Units. This subject relates to Sections 5.2.4 and 6.6 of the Ilidland SER. A list of mooting attendees is attached as Enclosure 1. is the agenda for the meeting and a conpilation of the handouts used in the course of the nceting, SUIMARY The purpose of the meeting was to obtain a more complete understanding of the PSI program.
It was pointed out that this issue is a joint responsibility between liRR and Region III, flRR will perform an evaluation of the examination sample and requests for relief from impractical examination requirements and Region III (K. Ward) will evaluate the examination procedures and the exanination results, fiRR's detailed review of the program is being conducted by Battelle PflL.
Issues discussed and agreements reached at this meeting are detailed belcw.
Bolting - Recent experience at the Oconee plant included a failure of thermal shield bolts.
In this instance 94 of 96 bolts fabricated of A-286 naterial were found to have experienced inservice degradation.
Consumers Power Company reported that all bolts used to fasten the thermal shield to the lower grid connection will be replaced with bolts fabricated of Inconel X750. A CPCo SAR change notice which would revise the FSAR to state this was distributed (see Enclosure 2).
It was reported that this information has been approved for incorporation into the FSAR but it was too late to be included in the next revision; the actual FSAR change incorporating this information will thus not be issued for approxinately 2 months.
CPCo was asked whether an analysis per flG 3133.6 of the thernal shield skirt would be provided. They responded that their consultant (i1PR) has expressed reservations abnut the appropriateness of this particular calculation. CPCo agreed to set up a discussion between their consultant and CDSellers (HTES) regarding this issue.
820429o 178 It was noted th3t preservice failure experience of high strength bolts and the techniques and procedures used in exanining belts were likely to be topics requiring discussion during ACRS consideration of the ti;Jland plant. These omce>,.
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e issues were discussed briefly and a copy of B&W procedure ISI-128, Rev. 2,
" Ultrasonic Examination of Studs and Bolting Greater than 2 Inches in Diameter," was offered for review by the staff (copies can be supplied upon request). Schedule and Status of PSI - CPCo reported that PSI piping examination in Unit 2 are 80% complete and in Unit I are 67% complete. A table showing the status of major examination efforts of this program was provided(seeEnclosure2).
Steam Generator Tube Inspection - The inspection of steam generator tubes included inspection of essentially the entire length of 100% of the tubes.
Multi-frequency testing equipment was used as necessary for clarification of single frequency results. The results of all testing are available on magnetic tape. There are approximately 6 tubes which were damaged during fabrication and will probably be plugged.
Effective Code Dates and Examination Sample - The preservice inspection of Class 1 piping welds included 100% of the Class 1 welds, except for a sampling process on CRD welds which is pemitted by Code or where examinations were detemined to be impractical and written relief is requested.
Examinations included the full length of longitudinal scan welds.
The Class 2 piping examination sample was based on Code Addenda through Summer 1975. The examination motheds were in accordance with Addenda through Sumer 1978. CPCo reported they had been quite strict in defining " parallel" lines for sample selection purposes and that this results in a conservative sample; substantial similarity in configuration as well as function was required for lines to be considered parallel.
CPCo agreed to provide, at NRC request, a concise sumary explaining the PSI program from a conceptual viewpoint. This summary is to include the editions used and why, the samples examined and how they were selected, a discussion of Code exclusions which were used and other information as discussed below.
Reactor Vessel Examination - The Code requires examination of " essentially 100%" of reactor vessel welds.
CPCo and SwRI described the extent of RPV examination and the examination techniques used. CPCo agreed to include in the summary discussed above a brief description of the examinations which were actually perfomed on the reactor vessel including near-surface limitations and the number and type of calibration blocks used.
It was noted that Regulatory Guide 1.150 had not been issued at the time the RPV examinations were performed. The "For Coment" version was available and was considered in establishing the scope of the examination. Ultrasonic signal gating of initial pulse effects was perfomed but the initial pulse was i
evaluated while perfoming the examination taking data and during evaluation.
I l
B&W Procedure ISI-133, Rev.1, " Ultrasonic Examination of Vessel Welds and l
Nozzle Inside Radius Sections," was provided for examination by the staff for l
information and available for review in the MTEB project file.
Examination of the High Pressure Injection Piping - Class 1 piping of 4 inch dianeter and less is-required by Code to receive a surface examination during PSI.
Class 2 lines of A inch diarmter and inss havn navor rnouirnd vnlunotric OFFICE )
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f or surface inspection during PSI in addition to the fabrication examination because of Section XI exemption criteria.
CPCo was infomed that the NRC disagrees with this position and strongly desires that Code exclusions not be used to eliminate all inspections on an Engine; red Safety Features line; a limited sampling of Class 2 welds is preferred. The core flood piping was cited as another example of where such an exclusion concern could arise.
The NRC position is that the ISI program should include a volumtric examination of the Class 1 and 2 piping in the core flood and high pressure injection systens. A sample would be appropriate; other applicants have used sample sizes of approximately 10%.
Relief Requests - The relief requests subnitted by CPCo are generally complete except for the following requests where additional infomation should be provided as indicated after the examination results have been evaluated by Consumers Power Company.
Request 11-12.
If the Staff accepts this request for relief, the Staff might require a change to the ISI scope at a later date if the " piggy back" mode of operation which foms the basis for sample selection becomes a frequent operating mode during plant operation.
Request 11-13. The NRC considers this is not necessarily an inconsistency in Code Requirements as indicated by the applicant.
It was also noted that a visual exaniration is being perfomed for other purposes (as part of hot functional walkdown) which could serve as an equivalent visual exam. CPCo will augment their discussion of alternative evaluations to indicate this exam.
(If applicable, this will also be done for requests II-15,16, and 17).
Request III-10. The relief request refers to an IWF visual examination of a reactor vessel support skirt weld after hot functional testing. CPCo agreed to augment their request by discussing examinations this weld received during fabrication, to discuss in more detail difficulties with removing vessel insulation, and to identify that other support skirt welds to be examined to IWF requirements.
In this regard, CPCo agreed to discuss their intentions with respect to IWF in the sumary document requested above.
Agreements Reached.
1.
The comitment to change in thermal shield bolt mterials have been adequately described by discussion in this meeting pending later issuance of an FSAR change per Enclosure 2.
The NRC is fomulating an overall generic position regarding a policy on bolting.
2.
As a part of the ISI review to be accomplished later, CPCo will consider adding reflectors to their calibration standard to improve capability to detect service induced problems in bolts.
3.
CPCo will provide, prior to SER issuance, a sumary document describing, in a qualitative manner sanple size criteria, offective Code dates, relief requests, and other infomation discussed above.
A copy of the sumary prepared by the Washington Public Power Supply System for WNP-2 was provided to CPCo as an example of the type of sumary desired (see
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The clarifications to relief requests discussed above will be submitted on a schedule to support being addressed in a Supplemental SER.
In' addition, a copy of the latest draft SER input for' Sections 5.2.4 and 6.6 was_ provided to CPCo for information (see E:1 closure 2).
R. II. Fernan, Project Manager Licensing Branch flo. 4 i
Division of Licensing i
Enclosures:
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MIDLAND Mr. J. W. Cook Vice President Consumers Power Conpany 1945 West Parnall Road Jackson, Michigan 49201 cc: Michael I. Miller, Esq.
Mr. Don van Farrowe, Chief Ronald G. Zamarin, Esq.
Division of Radiological Health Alan S. Farnell, Esq.
Department of Public Health Isham, Lincoln & Beale P.O. Box 33035 Suite 4200 Lansing, Michigan 48909-1 First National Plaza
- Chicago, Illinois 60603 William J. Scanlon, Esq.
2034 Pauline Boulevard James E. Brunner, Esq.
Ann Arbor, Michigan 48103 Consumers Power Company 212 West Michigan Avenue U.S. Nuclear Regulatory Comission Jackson, Michigan 49201 Resident Inspectors Office Route 7 Ms. Mary Sinclair Midland, Michigan 48640 5711 Summerset Drive Midland, Michigan 48640 Ms. Barbara Stamiris 5795 N. River Stewart H. Freeman Freeland, Michigan 48623 Assistant Attorney General State of Michigan Environmental Mr. Paul A. Perry, Secretary Protection Division Consumers Power Company 720 Law Building 212 W. Michigan Avenue Lansing, Michigan 48913 Jackson, Michigan 49201 Mr. Wendell Marshall Mr. Walt Apley Route 10 c/o Mr. Max Clausen Midland, Michigan 48640 Battelle Pacific North West Labs (PNWL)
Battelle Blvd.
Mr. Roger W. Huston SIGMA IV Building Suite 220 Richland, Washington 99352 7910 Woodmont Avenue Bethesda, Maryland 20814 Mr. I. Charak, Manager NRC Assistance Project Mr. R. B. Borsum Argonne National Laboratory Nuclear Power Generation Division
' lim South Cass Avenue Babcock & Wilcox
- pne, Illinois 60439 7910 Woodmont Avenue, Suite 220 l
Bethesda, Maryland 20814 mes G. Keppler, Regional Administrator S. Nuclear Regulatory Commission, 4
Cherry & Flynn Region III Suite 3700 799 Roosevelt Road Three First National Plaza Glen Ellyn, Illinois 60137 l
Chicago, Illinois 60602 l
Mr. Steve Gadler 2120 Carter Avenvy St. Paul, Minnesota 55108
ENCLOSURE 2 AGEllDA FOR MIDLAND,PLAliT, PRESERVICE 1riSPECTI0li PROGRAM
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Participants:
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f4RC - M. R. Hum, Materials Engineering Branch IAC Consultants - T. Taylor, Battelle Ptil G. Spanner, Battelle P!:L 1.
Objectives of PSI Program Review.
- 2.
Overview of Schedule and Status of PSI.
3.
Detailed Explanation of Effective Code Date's and Examination Sample
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4.
Reactor Vessel Examination - Procedures', Results, Areas that are Impractical to Examine, Reg. Guide 1.150.
5.
PSI of Steam Generator Tubes, Extent of Examination.
l' 6
Examination of Bolting.
7.
Ext.ination cf Class 2 Pcrtions.cf Mich Pressure :njection Pipinc.
5.
Reiief Requests.
'9.
tiRC Questions and Discussions.
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SUPPLEi.'El;TAL I*.EETit!G OUEST101:S Reactor Vessel Examination
. >.C (1) Discuss the techniques that will be used t'o detect near sur ace flaws
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on the. inner surface of the reactor vessel.. If electronic. gating of r-the ultrasonic signal will be used to suppress near-surface effects, estimate the depth of the inner surface that will not be examined.
(2), Identify the specific areas where the vessel welds could not be examined to the extent required by Section XI.
(3) Describe the design and access for examination of the reactor vessel support skirt weld.
Ecitino Examination (1)
Discuss U.T. procedures and calibration standards for bolting examina-tions.. Describe the f abrication and prese'rvice examinations of t.he reactor vessel thermal shield bolting.
Code Edition Selection (1)
Discuss the background and philosophy of your selection of various-Code Editions and Addenda.
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QUALITY l.SfAT ' 'JCE F ROGR AM T 0'
- 1. N FSAR O f 00
- 3. No. 7eo JOB NO.[7 2 O
- 2. DISCIPLINE / COMPANY
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- 4. ORIGINATOR f
- 6. REFERENCED SECTIONS OF SAR
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- 8. REFERENCED SPECIFICATIONS OR DRAWINGS NA S. JUSTIFICATION u
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I O. BECHTEL DISCIPLINE INTERFACE R2 VIEW:
INTERFACING STAFF REVIEW:
O ARCH O PLANT DSN
- d O ARCH O MECH O CIVIL O PQAE DCML O NUCLEAR O CONTROL SYS O STRESS O CONTROL SYSTEM O PLANT DSN O ELEC O OTHER O ELEC D REllABILITY 4 f e* a -- <d O GEOTECH O STRESS
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- 11. REVIEWED BY DATE
- 12. REVIEWED BY DATE
- 13. REVIEWED BY DATE (Group Supervisor)
(SAR COORDINATOR)
(NUCLEAR ENGINEER)
- 14. CONCURRENCE BY DATE
- 15. APPROVED BY (CPCo)
DATE
- 16. CONCURRENCE BY DATE (PROJECT ENGINEER)
(NSSS SUPPLIER)
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Summer 1978, and Section IX 'of the ASME Code and is monitored to 139 onsure compliance with the variables used in the weld procedure 33 qualification.
See Appendix 3A for a discussion of compliance with Regulatory Guide 1.31.
All parts of the CRDMs that are in contact with the reactor coolant or are part of the pressure boundary meet the rcquirements of B&W approved cleanness specifications.
These spscifications limit the level of contamination present on the surfaces that will contact reactor coolant.
Methods and materials used in cleaning and fabricating materials are restricted by these specifications and are used in accordance
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with written procedures.
All cleaning is performed in suitable environments for these operations.
After final cleaning, cleanliness is maintained by sealing or wrapping in selected matorial, such as polyethylene.
Onsite storage and cleanness controls are specified by appropriate field specifications.
These cleanliness and contamination controls on the CRDMs meet the intent of Regulatory Guide 1.37.
See Appendix 3A for a 33 discussion of compliance with Regulatory Guide 1.37 4.5.2 REACTOR INTERNALS MATERIAL 4.5.2.1 Material Soecifications 4%j w Ajfeactor core support assembly (CSA) components are fabricated primarily from Type 304 stainless steel (i.e., plate, A-240 Typa 304; forgings, A-473-63 Type 304; tubes, A-269 or 312 Type 304L; bar, A-276 Type 304). -The reactor core support assembly (CSA) consists of three main subassemblies:
plenum assembly (including upper grid assembly, plenum cover, column-weldments, plenum cylinder, and control rod guide structure),
core support shield assembly and the lower core support assembly 8
l (including thermal shield, core barrel assembly, lower grid, flow l
distributor, and guide tube assembly).
Therefore reference to the reactor core support assembly includes components as listed in Section 4.5.2 of Standard Review Plan.
Materials and weld metal comply with the applicable requirements of the ASME Code.
Cold worked austenitic stainless steel having yield strength greater than 90,000 psi have not been used in the fabrication of the CSA.
The adequacy and suitability of the material are discussed in Subsections 5.2.3.2.2 and 5.3.3.2.
The reactor core l8 support assembly component weld metals are ASME II(c) SFA-5.9 33
'ER308L) or SFA-5.4 (E308).
ASTM B 304 (ERNiCr-3 ) was used for minor attachments such as dowels.
l 8 1rSCRT'A 4.5.2.2 Cont 2 s on Welding The controls on welding of austenitic stainless steels are discussed in Subsections 5.2.3.4 and 5.3.1.4.
Revision 39 4.5-2 11/01
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MA1ERIALS EfJGifEERIfjG BRAf!CH IfJSERVICE If!SPECTICf1 SEC110N 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing This section was prepared with the technical assistance of DOE contractors from the Pacific Northwest Laboratories.
5.2.4.1 Compliance with the Standard Review Plans The July l98L 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 Ccolant Pressure Boundary Inservice Inspection and Testing".
Our review is continuing because the i
applicant has not completed all preservice examinations.
Our 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 reviewed because this area applies only to inservice inspection (ISI), not to the PSI.
This subject will be addressed during review of the ISI program after licensing.
Paragraph 11.5, " Acceptance Criteria, Evaluation of Examination R e su lt s'> r has been reviewed and the applicant has incorporated ASME Code Section IWB-3000, " Standards for Examination Evaluation" into his PSI Program. However, ongoing NRC generic activities Note:
() indicates sentences that will be added upon acceptable resolution.
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9 maximum acceptable size flaws specified in the IWB-3000 acceptance standards, for exampler ASME Code procedures specified for volumetric examination of reactor vessels, bolts and studs, and piping have not proven to be capable of detecting the maximum acceptable size flaws We will continue to evaluate development of improved in all cases.
procedures and will require that these improved procedures be made We have not a part of the inservice examination requirements.
reviewed the applicant's repair procedures based on ASME Code Section IWB-4000r " Repair Procedures" Repairs are not. generally necessary in This subject will be addressed during our review of the PSI program.
the ISI program.
Paragraph 11.8, " Acceptance Criteriar Relief Requests"r has not been completed because the applicant has not identified all limitations to examination. Specific areas where ASME Code examination requirements cannot be met wiLL be identified as performance of the PSI progresses.
Our complete evaluation of the PSI program will be presented in a supplement to this Safety Evaluation Report af ter the applicant submits the required examination information and identifies all plant-specific areas where ASME Code Section XI requirements cannot be met and provides supporting technical justification.
5.2.4.2 Examination Requirements k
General Design Criterion 32, " Inspection of Reactor Coolant Pressure Doundary", Appendix A of 10 CFR Part 50 requires, in part, that
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'n-be desi ined 1o per mit periodic inspection and iesting of in,por t ant -
t areas and f eatures to assess their structural and leak-tight integrity.
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To ensure that no deleterious defects develop during service, selected welds and weld-heat affected-zones (HAZ) will be inspected periodically at Midland 1 and 2.
O The design of the ASME Code Class 1 and 2 components of the reactor coolant pressure boundary incorporate provisions for access for inservice inspections, as required by Paragraph IWA-L500 of Section XI of the ASME Code.
Section 50.55a(g),10 CFR Part 50, defines the detailed requirements for the preservice and inservice programs for light water cooled nuclear power facility components.
Based upon the construction permit date of December 15, 1972, this section of the regulations requires that a preservice inspection program be developed and implemented using at least the Edition and Agenda of Section XI of the ASME Code in ef f ect six months prior to the date of issuance of the construction permit.
Also, the initial ISI program must comply with the requirements of the latest Edition and Agenda 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.
w 5.2.4.3 Evaluation of Compliance with 10 CFR 50.55a(g)
We have reviewed the information in the FSAR and the Preservice Inspection Plan through Revision 1 dated October 30, 1981 and received with a letter dated November 19, 1981. The preservice examination was discussed at a andatthattime,theapplicantprovidedf public meeting on March 18, 1981,
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.o the appticant's Preservice Inspection Program to be acceptable.]
The applicant has performed approximately 80 percent of the preservice examination which is based on the 1977 Edition of Section XI of the ASME Code, including Agenda through Summer 1978.
This represents a voluntary updating by the applicant f rom the 1971 Edition of Section XI including Agenda through Winter 1971 required by the regulation. The applicant has incorporated into the PSI Program requirements f rom other editions of Section XI referenced in 10 CFR 50.55a(b) as permitted by the regulation. The most significant alternative provisions that the applicant is using f rom other code editions are listed as follows:
1.
The Unit 2 pressurizer and steam generators were examined in 1974 Edition of the Code while they were still at the fabrication shop.
2.
Class 1 pipe branch connection welds will be examined to the 1980 Edition, Winter 1980 Agenda.
Figures IWB-2500-9.10 and 11 more thoroughly define the weld volume subject to inspection in the Winter 80 addendum.
3.
The 1977 Edition through Summer 1979 Addenda will be used for l
examination evaluation of planar flaws. These ruled clarify and supersede those of the 1977 Edition through the Summer 1978 Addenda.
l l
4.
Class 1, 2, and 3 Component Supports will be examined per the rules of Subsections IWA and IWF included in the 1977 Edition through tFe Summer 1979 Addenda.
5.
In accordance with Regulatory Guide 1.147, (February 1981 revision) the applicable deviations specified in Code Case N-216 (" Alternative Rules for Reactor Vessel Closure Stad examination") will be utilized.
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m 1978 Addenda of Sect ion XI and 1ind the except ion:, to t>e accept able alternatives since they are based on editions of the Code referenced by 10 CFR 50.55a(b).
The applicant has identified several limitations to examination. Of particularsignificancearesoneportionsofthereactorvessel.(We 3
have requested a detailed listing of such areas. The applicant has e
committed to identify all impractical ASME Code examination requirements with a supporting technical justification.
We will complete our evaluation of requests for relief f rom impractical examination requirements in a supplement to this SER.
The initial inservice inspection program has not been submitted by the applicant. We will evaluate the program after the applicable ASME Code Edition and Addenda can be determined based on Section 50.55a(b) of 10 CFR Part 50, but before the first refueling outage when ISI commences.
5.2.4.4 Conclusions The conduct of periodic inspections and hydrostatic testing of pressure retaining components of the reactor coolant pressure b ounda ry, in accordance ith the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and 10 CFR Part 50, will provide reasonable assurance that evidence of structural degradation or loss of leaktight integrity occurring during service will be detected in time to permit corrective action before the safety functions of a component are compromised.
inspection; required by Compliance with the preservice and iriservice 50 constitutes an acceptable basis for the Code and 10 CFR Part satisf ying the inspection requirements of Criterion 32 of the General Design Criteria.
s 5.2.4.5 References 1.
NUREG-0800r Standard Review Planse Section 5.2.4r " Reactor Coolant Boundary Inservice Inspection and Testing," July 1981.
Code of Federal Regulationse Volume 10, Part 50.
2.
American Society of Mechanical Engineers Boiler and Pressure 3.
Vesset Coder Section XI 1971 Editioni through Winter 1971 Addenda B
1974 Edition, through Summer 1975 Addenda 1977 Editions through Summer 1979 Addenda g
1980 Editioni through Winter 1980 Addenda ti e
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Inservice Inspection of Class 2 and 3 Components 2
6.6 This section was prepared with the technical assistance of DOE contractors from the Pacific Northwest Laboratories.
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6.6.1 Compliance with the Standard Review Plans The July 1981 Fdition of the " Standard Review Plan for the Review (SRPr of Safety Analysis Reports for Nuclear Power Plantsr" l
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fJURLG-0800) ine tudes Sect ion 6.6 " Inservice Inspect ion of Class 2 and 3 Components." Our review is continuing because the applicant has not completed all preservice examinations.
Our review to date was conducted in accordance with Standard Review Plan Section 6.6 except as discussed below.
1 Paragraph II.4, " Acceptance Criteriar Inspection Intervals," has c
not been reviewed because this area applies only to 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 and the applicant has incorporated ASME Code Sections IWC-3000 and IWD-3000, " Standards for Examination Evaluation" into his PSI program.
However, ongoing NRC generic activities and research projects indicate that the presently specified minimum ASME Code procedures may not always be capable of detecting the maximum 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 maximum acceptable size flaws in all cases. We wiLL continue to evaluate development of improved procedures and will require that these improved procedures be made a part of the inservice examination requirements.
We have not r..,
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.' t r eviewed the applicant 's repair pr ocedures based on ASME Code
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Sections IWC-4000 and IWD-4000, " Repair Procedures." Repairs are not generally necessary in the PSI program.
This subject wiLL be addressed during our review of the ISI program.
i Paragraph II.9, " Acceptance Criteriar Relief Requests," has not been completed because the applicant has not identified the limit ations to examination.
Specific areas where ASME Code examination requirements cannot be met will be identified as performance of the PSI progresses.
Our complete evaluation of the PSI program will be presented in a supplement to this 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 supporting technical justification.
6.6.2 Examination Requirements General Design Criteria 36, 39, 42, and 45, Appendix A of 10 CFR Part 50, requirer 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 PSI programs for light water cooled nuclear power facility components.
Based upon the c ont.t ruct i on permit date of D e c,enibe r 15, 1972, thit, section of the regulat ions requires that a PSI program for Class 2 and 3 component s be developed and implemented using at -least the Edition and Addenda of Section XI of the ASME Code in effect six months prior to the date of issuance of the construction permit.
Also, the initial Inservice Inspection program must comply with the requirements of the latest Edition and Addenda of Section XI of the ASME Code in ef fect twelve months prior to the date of issuance of the operating licenser 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)
We have reviewed the information in the FSAR and the Preservice Inspection Plan through Revision 1 dated October 30, 1981 and received with a letter dated November 19, 1981. The preservice examination was discussed at a public meeting on March 18, 1982, and at that time the applicant also provided supplemental information2}(Basedontheaboveinformation,wefindtheapplicant's Preservice Inspection Program to be acceptable.)
The applicant has performed approximately 80 percent of the preservice examination which is based on the 1977 Edition of Section XI of the ASME Coder including Addenda through Summer 1978.
This represents a voluntary updating by the applicant from the 1971 Edition of Section XI including Addenda through Winter 1971 required
by the regulat ion.
The applicant has incorporated int o the PSI program requirements from later editions of Section XI Code referenced in 10 CFR 50.55a(b) as permitted by the regulation.
The most significant alternative provisions that the applicant is using for ASME Code Class 2 components is discussed in SER Sect ion 5.2.4.3.
We have reviewed these exceptions to the 1977 Edition, Summer 1978 Addenda of Section XI and find the exceptions
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to be acceptable alternatives since they are based on editions referenced by 10 CFR 50.55a(b).
The applicant has committed to identify all plant-specific areas where the Code requirements cannot be met and provide supporting technical justification for relief. We will complete our evaluation of requests for relief f rom impractical examination requirements in a supplement to this SER.
The initial' inservice inspection program has not been submitted by the applicant.
We will evaluate the program after the applicable ASME Code Edition and Addenda can be determined based on Section 50.55a(b)
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of 10 CFR Part 50s but before the first refueling outage when ISI
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Commences.
In addition to generating electricity, the Midland plant also supplies process steam to a chemical plant located adjacent to the plant site.
The process steam is produced in a tertiary heat exchanger.
Un i t '_1 is the primary supplier of heat to produce the process steams asthough an intertie line between units 1 and 2 is provided so that unit 2 can also supply the heat if unit 1 is down.
All of the
6 w
tompnoentL ar nf piping invotved in produt ing ihe ter 1ior) p r o c e s:.
steam are out side the scope of Section XI requirement s and theref ore have no effect on this SER.
6.6.4 Conclusions Compliance with the preservice and inservice inspections.equired by the American Society of Mechanical Engineers 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 Plani Section 6.6, " Inservice Inspection of Class 2 and 3 Components," July 1981.
2.
Code of Federal Regulations, Volume 10, Part 50.
.1 3.
American Society of Mechanical Engineers Boiler and Pressure J
Vessel Coder Section XI 1971 Editioni through Winter 1971 Addenda 1974 Editions through Summer 1975 Addenda 1977 Edition, through Summer 1979 Addenda 1980 Editions through Winter 1980 Addenda 1
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ENCLOSURE 1 a
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LIST OF ATTENDEES s.
March 15,1982 NRC Consumers Power Co. (CPCo)
H. Slager M. Hum R. Huston C. Sellers T. Hollowell
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.t D. Smi th G. Eichenberger C. Cheng R. Hernan Southwest Research Institute (SwkI)
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PNL INRC consultant)
W. Flach T. Taylor 1,
General Physics Corp.
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SUMMARY
DISTRIBUTION Docket File G. Lear NRC/PDR S. Pawlicki Local PDR V. Benaroya TIC /NSIC/ TERA Z. Rosztoczy LB #4 r/f W. Haass H. Denton D. Muller E. Case R. Ballard D. Eiser. hut W. Regan R. Purple R. Mattson B. J. Youngblood P. Check 4)
Rv A. Schwencer
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APR 081982 m. ??
R. Vollmer J. Hulman 6
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a nna gy,,@ 8 J. P. Knight W. Gammill na_n.e R. Bosnak L. Rubenste F. Schauer T. Speis e
5' R. E. Jackson W. Johnston cn Attorney, OELD l S. Hanauer OIE y
C. Berlinger ACRS (16)
F. Schroeder R. Tedesco D. Skovholt D. Hood M. Ernst 2
K. Kniel NRC
Participants:
G. Knighton M. Hum A. Thadani C. Sellers D. Tondi D. Smith J. Kramer C. Cheng D. Vassallo R. Hernan P. Collins D. Ziemann F. Congel J. Stolz M. Srinivasan R. Baer E. Adensam R. Hernan Project Manager bec:
Applicant & Service List Licensing Assistant a. uuncan l
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