ML20246D524
| ML20246D524 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/07/1989 |
| From: | Reynolds S Office of Nuclear Reactor Regulation |
| To: | Firlit J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| TAC-71843, NUDOCS 8907110356 | |
| Download: ML20246D524 (9) | |
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July 7,1989 F-Docket No. 50-312 Mr. Joe Firlit Chief Executive Officer, Nuclear Rancho Seco Nuclear Generating Station 14440 Twin Cities Road Herald,. California 95638-9799 l
Dear Mr. Firlit
SUBJECT:
SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN, REQUEST FOR ADDITIONAL INFORMATION (TAC NO. 71843)
We are reviewing and evaluating the Second 10-Year Interval Inservice Inspection Program Plan for the Rancho Seco Nuclear Generating Station.
Additional information is needed for completion of our review and evaluation. We request that you provide a response to the attached list of questions within forty-five (45) days of receipt in order to meet our projected schedule.
Should you have any questions regarding this request, please communicate with the NRC project manager assigned to your plant.
Sincerely,
/s/
Steven A. Reynolds, Project Manager Project Directorate V Division of Reactors Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
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July 7, 1989 Docket No. 50-312 Mr. Joe Firlit Chief Executive Officer, Nuclear Rancho Seco Nuclear Generating Station 14440 Twin Cities Road Herald, California 95638-9799 Dear Mr. Firlit
SUBJECT:
SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN, REQUEST FOR ADDITIONAL INFORMATION (TAC NO. 71843)
We are reviewing and evaluating the Second 10-Year Interval Inservice i
Inspection Program Plan for the Rancho Seco Nuclear Generating Station.
Additional information is needed for completion of our review and questions within forty-five (you provide a response to the attached list of evaluation. We request that
- 45) days of receipt in order to meet our projected schedule.
Should you have any questions regarding this request, please communicate with the NRC project manager assigned to your plant.
Sincerely, 2L a A Steven A. Reynol s, Project Manager Project Directorate V Division of L'eactors Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
As st;ted cc: See next page l
__-,,---an_a--------_-,----_---..-----,-,..u-u---
i Mr. Joe Firlit Rancho Seco Nuclear Generating Station cc:.
Mr. - David S. Kaplan, Secretary Mr. John Bartus and General Counsel Ms. JoAnne Scott Sacramento Municipal Utility District Federal Energy Regulatory Commission 6201 S Street 825 North Capitol Street, N. E.
P.O. Box 15830 Washington, D.C.
20425 Sacramento, California 95813 ThomasAiBaxter,Esq.
Ms.-Helen Hubbard Shaw, Pittman, Potts & Trowbridge P. O. Box 63 2300 N Street, N.W.
Sunol, California 94586 Washington, D.C.
20037 Mr. Steven Crunk Manager, Nuclear Licensing Sacrahento Municipal Utility. District l
Rancho Seco Nuclear Generating Station 14440 Twin Cities Road l
Herald, California 95638-9799 l
Mr. Robert B. Borsum, Licensing Representative Babcock & Wilcox Nuclear Power Division i
l 1700 Rockville Pike - Suite 525 Rockville, Maryland 20852 Resident Inspector / Rancho Seco c/o U. S. N. R. C.
14440 Twin Cities Road Herald, California 95638 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Dr. Gerard C. Wong, Chief Radiological Materials Control Section State Department of Health Services 714 P Street, Office Building #8 Sacramento, California 95814 Sacranento County Board of Supervisors 700 H Street, Suite 2450 Sacramento, California 95814 (10)
e SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NUMBER 50-312 MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEERING AND SYSTEMS TECHNOLOGY Recuest for Additional Information - Second 10-Year Interval Inservice Inspection Procram Plan 1.
Scoce/ 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) which are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except design and access provisions and preservice examination requirements, set forth in the 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 intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein.
The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code which are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein. Based on the start date of September 18, 1988 for the second 10-year interval, the Inservice Inspection (ISI) Program Plan has been prepared to meet the requirements of the 1986 Edition of ASME Code Section XI.
Asrequiredby10CFR50.55a(g)(5),iftheLicenseedeterminesthat.
certain Code examination requirements are impractical and relief is requested, the Licensee shall submit information to the Nuclear i
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Regulatory Commission (NRC) to support that determination.
t The staff has reviewed the available information in the Rancho Seco Nuclear Generating Station Second 10-Year Interval ISI Program Plan, submitted December 30, 1988, and the requests for relief from the ASME Code Section XI requirements which the Licensee has determined to be impractical.
2.
Additional Information/ Clarification Reauired Based on the above review, the staff has concluded that the following information and/or clarification is required in order to complete the review of the ISI Program Plan:
A.
Provide a listing of all ASME Nuclear Component Code Cases being used during second 10-year interval ISI examinations at Rancho Seco.
B.
Augmented examinations have been established by the NRC when added assurance of structural reliability is deemed necessa'y. Examples r
of documents which may require augmented examination are:
(1) High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, Branch Technical Position ASB 3-1; (2)
Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations; (3) Regulatory Guide 1.14, " Reactor Coolant Pump Flywheel Integrity;" and Address these and any other augmented examination which may have been incorporated in the Rancho Seco Nuclear Generating Station Second 10-Year Interval Inservice Inspection Program Plan.
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t C.
Provide a listing of all Class 2 Residual Heat Removal (RHR),
Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) systems at Rancho Seco and include the total number of welds in each of these systems.
Staff review of all Class 2 piping welds receiving volumetric examinations during the second 10 year inspection interval at Rancho Seco shows the following:
Pipe Sites >4" NPS and Wall Thicknesses 13/8" Volumetric System and Surface Aux. FW 5
Decay Heat A 5
Decay Heat B J
Total Welds 13 Pipe Sizes 12" and 54" NPS and Wall Thickness >1/5" Volumetric System and Surface Make Up Discharge 8
HPI A Discharge 7
HPI B Discharge 8
HPI Mini Flow J
Total Welds 26 A representative sampling of welds in the RHR, ECC, and CHR systems should receive inservice volumetric examinations. The staff has previously determined that a 7.5% augmented volumetric sample constitutes an acceptable resolution at similar' plants. Discuss the impact of performing volumetric examination of at least a 7.5%
sampling of the Class 2 piping welds in these systems.
D.
Review of Section 1 of the ISI Plan listing the NDE examinations being performed during the second interval and the calibration block drawings in Section 7 shows that some of the calibration blocks may not meet the applicable Code requirements. Examples are as follows: Calibration Block #26 (10-inch diameter, 1.125-inch thick) is being used for ISI examinations of Item B09.11-27 (12.8-inch diameter,1.3-inch wall thickness) and Item B09.11-31 j
(14-inch diameter,1.4-inch wall thickness).
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Calibration Block #27 (3-inch thick flat block) is being used for
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the examination of Items B09.31-1 thru B09.31-4 (small diameter branch connections-to-large diameter primary coolant system piping). Calibration Block #23 (stainless steel) is being used to examine Item B05.040-4 (carbon steel-to-inconel dissimilar metal
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weld).
i It is also noted that many of the calibration block drawings in
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Section 7 of the Plan have been reduced in size :nd are illegible with regard to dimensions and material specifications.
Appendix III, " Ultrasonic Examination of Piping Systems," of Section XI of the Code requires that basic calibration blocks be-made from material of the same nominal diameter and nominal wall thickness or pipe schedule as the pipe to be examined. The calibration blocks for similar metal welds shall be fabricated from the material specified for the piping being joined by the weld.
Calibration blocks for dissimilar metal welds shall be fabricated from the material specified for the side of the weld from which the examination will be conductek if the examination will be conducted from both sides, calibration reflectors shall be provided in both materials.
The staff considers inservice volumetric examinations of Code Class I and 2 systems cr9tial to plant safety and, therefore, feels that proper calibration standards should be obtained and utilized for all ISI examinations.
Provide a discussion of the calibration blocks being used for ISI examinations during the second 10-year interval at Rancho Seco and either confirm that all calibration blocks meet or exceed the Code requirements or provide technical justifications in the form of requests for relief for the continued use of any blocks which do not meet the Code requirements.
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E.
- Relief Reatest #2: Relief is requested from the ASME Code-required s.
surface enmination of RPV core flood nozzle safe end welds. The Licensee has proposed performing a volumetric examination of 100%
of-the pipe thickness with automated inspection equipment from the nozzle I.D.
The proposal could be considered acceptable provided that the Licensee meets the following conditions:
(1) The remote volumetric examination includes the entire weld volume and heat affected zone ins'tead of only the inner one-third of the weld as required by the Code.
(2) The ultrasonic testing instrumentation and procedure are demonstrated to be capable of detecting 0.D. surface-connected defects, in the circumferential orientation, in a laboratory test block. The defects should be cracks and not machined
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notches.
Provide a discussion of the above conditions and verify that they will be met.
F.
Relief Request'#3: Relief is requested from performing the ASME.
Code-required volumetric examination of Primary Coolant Pump casing welds and visual (VT-3) examination of the pump casing internal surfaces. The Licensee has proposed performing a visual examination of 100% of the external surfaces of the welds in lieu of the Code-required volumetric examination.
Other plants with similar pump configurations have committed to performing surface examinations of the exterier surfaces of the welds once per inspection interval and, if the pumps are disassembled for maintenance, a surface examination of the accessible interior surfaces of the welds. Discuss the impact of performing sn'fs ce examinations as described above in lieu of the s
proposed visuai 3.xamination for these welds.
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G.
Relief Request #4 (Class 2 hydrostatic test) and Relief Request f 5 (Class 3 hydrostatic test): Discuss the operating and design pressures of the affected components as compared to the Code-required hydrostatic test pressure. As it is noted that the proposed substitute examination (a leak check during normal system operation) is a Code requirement and not a substitute examination,.
include a discussion of the design pressure of the affected pump seals giving consideration as to what the maximum alternative test pressure could be in order to meet the intent of the Code.
The schedule for timely completion of this review requires that the Licensee provide, by the requested date, the above requested information and/or clarifications with regard to the Rancho Seco Nuclear Generating Station Second 10-Year Interval ISI Program Plan.
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