ML20057B600
| ML20057B600 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/16/1993 |
| From: | Olshan L Office of Nuclear Reactor Regulation |
| To: | Donnelly P CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TAC-M85744, NUDOCS 9309220304 | |
| Download: ML20057B600 (7) | |
Text
Docket No. 50-155 DISTIiINEN '
' Docket: File &
BJorgensen, RIII NRC & LPDRs L01shan Mr. Patrick M. Donnelly, Plant Manager PD31 Rdg File Big Rock Point Plant JRoe Consumers Power Company JZwolinski 10269 U.S. 31 North WDean Charlevoix, Michigan 49201 CJamerson OGC
Dear Mr. Donnelly:
ACRS(10)
SUBJECT:
BIG ROCK POINT PLANT - REQUEST FOR ADDITIONAL INFORMATION - THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF (TAC NO. M85744)
The staff, with assistance from its contractor, Idaho National Engineering Laboratory (INEL), is reviewing and evaluating the Third 10-Year Interval Inservice Inspection Program Plan and requests for relief from the ASME Boiler
& Pressure Vessel Code,Section XI requirements submitted January 22, 1993, for Big Rock Point Plant. Additional information is required in order for us to complete our review.
Please provide a response to the enclosed request for additional information within 60 days of receipt of this letter.
To expedite the review process, please also provide a copy of the response to our contractor, INEL, at the following address:
Boyd W. Brown EG&G Idaho, Inc.
INEL Research Center 2151 North Boulevard PO Box 1625 Idaho Falls, Idaho 83415-2209 The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, Original sigrei by Leonard N. Olshan, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/ enclosure:
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Mr. Patrick M. Donnelly, Plant Manager Big Rock Point Nuclear Plant cc:
Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Jane E. Brannon, County Clerk County Building Annex 203 Antrim Street Charlevoix, Michigan 49720 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. O. Box 30195 Lansing, Michigan 48909 U.S. Nuclear. Regulatory Commission Resident Inspector Office Big Rock Point Plant 10253 U.S. 31 North Charlevoix, Michigan 49720 Mr. David P. Hoffman, Vice President Nuclear Operations Big Rock Point Plant Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 w nm
I Enclosure CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NUMBER 50-155 Reouest for Additional Information - Third 10-Year Interval Inservice Inspection Procram Plan 1.
Scope / Status of Review Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class I, Class 2, and Class 3 meet the requirements, except design and access provisions and.preservice examination requirements, set forth in ASME Code Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during the successive 120-month inspection interval comply with the requirements in the latest edition and addenda of the ASME Code incorporated by the reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of a successive 120-month interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of the ASME Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein.
The licensee, Consumers Power Company, has prepared the Big Rock Point Plant, Third 10-Year Interval Inservice Inspection (ISI) Program Plan, Revision 0, to meet the requirements of the 1986 Edition of Section XI of the ASME Code. As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain ASME Code examination requirements are impractical and requests relief, the licensee shall submit information to the Nuclear Regulatory Commission (NRC) to support that determination.
The staff has reviewed the available information in the Big Rock Point Plant, Third 10-Year Interval ISI Program Plan, Revision 0, submitted January 22, 1993, which included the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical.
2.
Additional Infomation Reovired Based on the above review, the staff has concluded that the following information is required to complete the review of the ISI Program Plan:
A.
Augmented examinations have been established by the NRC when added assurance of structural reliability is deemed necessary.
Examples of dccuments that address augmented examinations are:
i i ;
(1) Branch Technical Position MEB 3-1, High Energy Fluid Systems, l
Protection Against Postulated Piping Failures in Fluid Systems Outside Containment; (2) Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations; (3) NUREG-0619, BWR Feedwater Nozzle and CRD Return Line Nozzle Cracking; and (4) NUREG-0803, Integrity of BWR Scram System Piping.
l Address the degree of compliance with these and any other augmented l
examination requirements that may have been incorp' orated in the Big Rock Point Plant, Third 10-Year Interval Inservice Inspection Program l
Plan.
B.
Effective September 8,1992, new regulations were issued 5
regarding the augmented examination of reactor vessels. As a result of these regulations, all licensees must augment their reactor vessel examinations by performing once, during the j
inservice inspection interval in effect on September 8,1992, I
the examinations required for reactor vessel shell welds specified in Item Bl.10 of Examination Category B-A of the 1989 ASME Code.
In addition, all previously granted relief l
for Examination Category B-A, Item Bl.10, for the interval in effect on September 8,1992, is revoked by the new regulation.
l Please provide the staff with the projected schedule and a technical discussion describing how the regulation will be implemented for these welds at the Big Rock Point Plant during the third 10-year interval.
Include in the discussion a description of the intended approach and any specialized techniques or equipment that will be used to complete the required augmented examination.
C.
Paragraph 10 CFR 50.55a(b)(2)(iv) requires that ASME Code Class 2 piping welds in the residual heat removal (RHR), emergency core cooling (ECC), and containment heat removal (CHR) systems be examined.
These systems should not be completely exempted from inservice volumetric examination based on Section XI exclusion criteria contained in Table IWC-2500-1. The staff has previously determined that a 7.5% augmented volumetric sample constitutes an acceptable resolution at similar plants. The Big Rock Point Plant is not of the conventional boiling water reactor design, and some system nomenclatures are not standard ASME Code terminologies. The staff requests that the licensee provide a cross reference listing of l
l systems or portions of systems that provide functions equivalent to RHR, ECCS, and CHR systems.
Include a listing of any welds selected to augment the inservice inspection program.
D.
Request for' Relief RR-A2, Vessel Bottom Head Meridional Welds, addresses the inaccessibility of the reactor vessel lower head I
meridional welds.
It appears that a portion of the meridional weld (s) may be accessible for examination in conjunction with the lower head circumferential, dollar plate weld 793-1.
Please provide an estimate of the percentage of coverage obtainable for the ASME Code-required examination area and provide a detailed sketch of the limitations associated with the circumferential weld and associated meridional welds.
E.
Request for Relief RR-A3, Primary Nozzle-to-Vessel Welds, Nozzle Inside Radius Sections, and Nozzle-to-Safe End Welds, addresses relief from all examinations associated with the 20-inch recirculation nozzles, except for accessible portions of the nozzle-safe end (B-F) welds and the 8-inch shutdown unloading nozzle 795-15.
It appears that nozzle-to-shell welds (796-IA and 796-18) are accessible from the outside surface only, with approximately one-third of the ASME Code-required examination area accessible for examination.
Please provide an estimate of the percentage of coverage obtainable for the ASME Code-required examination area. Provide a detailed sketch of the nozzle-to-shell weld area and associated access limitations.
It is stated that a portion of the nozzle-safe end Irelds on the 20-inch recirculation nozzles is accessible and will be examined.
Please provide an estimate of the percentage of coverage obtainable for the areas requiring volumetric and surface examinations. Relief is requested for all examinations associated with nozzle 795-15.
Provide a detailed sketch of the nozzle-to-shell and nozzle-to-safe end and their associated access limitations.
The ASME Code requires that examinations be performed to the extent practical. Advanced nondestructive examination techniques and equipment have been developed that allow examination of areas that previously required relief (e.g.,
inner radius sections, limited clearances). Describe any advanced examination techniques that may permit examination of portions of areas for which relief has been requested.
F.
Request for Relief RR-A4, 3-Inch Reactor Vessel Nozzles, addresses relief for nozzle-to-shell, nozzle inside radius, and dissimilar metal welds.
Provide a detailed sketch of the subject nozzle examination areas showing the access limitations. As stated, nozzle-to-shell welds 795-1C, -ID, and -lE are accessible for examination. Describe the technique used to perform the ASME Code-required examination.
In regard to the discussion of the unavailability of calibration block materials, the staff continues to monitor the development of new or improved examination techniques, including examination area mockups. Calibration block material al.ternatives should be explored for materials with equivalent ultrasonic properties. As improvements in these areas are achieved, the staff is requiring that the new
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techniques be made part of the ISI program. Discuss reviews of new and improved examination techniques that may be incorporated into the Third 10-Year Interval Inservice Inspection Program.
G.
Request for Relief RR-A5 addresses relief from the ASME Code-required surface examinations on the 14-inch steam outlet nozzles 795-11a to -Ilf. The licensee proposes a mechanized ultrasonic examination of accessible portions in lieu of the ASME Code-required surface examination. This proposal could be considered acceptable if the following conditions were met:
(1) The remote volumetric examination includes the entire weld volume and heat-affected zone instead of only the inner one-third of the weld as in ASME Code-required volumetric examinations.
(2) The ultrasonic testing instrumentation and procedures are demonstrated to be capable of 1
detecting 0.D. surface-connected flaws in laboratory test blocks. The laboratory test blocks should contain crack-like defects and not machined notches.
Please provide a discussion of the above conditions and verify that they will be met.
H.
Relief Requests RR-A6 and RR-A7 address relief associated with the steam drum.
The steam drum appears to be an extension of the reactor pressure vessel that provides steam separation similar to a steam separator in a conventional boiling water reactor; the steam drum does not fall into any of the designated ASME Code examination categories. Please provide detailed information regarding the steam drum design.
In addition, describe the insulation type, access, and radiation levels associated with the steam drum when drained for refueling. The steam drum has manways at each end of the vessel. Describe the accessibility and the feasibility for performing ASME Code-required examinations remotely from the vessel interior.
i I.
Requests for Relief RR-All and RR-A12 appear to address (1) volumetric examinations associated with the cleanup and main recirculation pump casing welds and, (2) visual examinations of the main recirculation pump internal surfaces.
The discussion of the basis for relief appears to apply only to the recirculation pump. Are the pump functions, designs, access, and exposures the same for the recirculation pumps and the clean-up pumps? Please provide clarification.
n-
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J.
Verify that there are no relief requests in addition to those submitted.
If additional relief requests are required, the i
licensee should submit them at this time for staff review.
I I
The schedule for timely completion of this review requires that the licensee i
provide, by the requested date, the above requested information and/or clarification with regard to the Big Rock Point Plant, Third 10-Year Interval ISI Program Plan.
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