ML20073S762
| ML20073S762 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/01/1994 |
| From: | Marsh W SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9407060157 | |
| Download: ML20073S762 (5) | |
Text
{{#Wiki_filter:_ 4 _f"d3? Southern California Edison Company 23 t'AFtKEFt ST F4E ET IAVINIT calif OnNIA 92710 MANaot.ML L M tot out A Y Ar t Asses (F1 ) 4 03 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 205S5 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 Second Ten Year Inservice Inspection Submittal San Onofre Nuclear Generating Station Units 2 and 3
Reference:
Letter from W. C. Marsh to the Document Control Desk of the U. S. Nuclear Regulatory Commission dated October 4,1993;
Subject:
ASME Code Update for the Second Ten year Interval, Inservice Inspection Programs, San Onofre Nuclear Generating Station, Units 2 and 3 This letter provides a correction to the San Onofre Units 2 and 3 Inservice Inspection (ISI) programs submitted for the second ten year interval which were submitted to the NRC by the referenced October 4,1993, letter. This also clarifies the statement included in that submittal regarding the reactor vessel inspections performed during the'first ten year inspection interval at San Onofre Units 2 and 3. Southern California Edison (Edison) determined the need to provide this correction while p(g)p(6)(ii)(A)(2). re aring answers for the NRC questionnaire concerning 10 CFR 50.55a UNITS 2 AND 3 ISI PROGRAM REVISION Paragraph 2.11.1 " Reactor Vessel, Category B_-A," found on page 22 of the ISI programs for both Units 2 and 3, submitted by Reference 1, states: "The au mented examination requirements in 10 CFR 50.55a( )(6)(li)(A)(2),1993 Revision of the Code of Federal Regulat ons Title 10. Part 50 were satisfied during the last period of. the first inspection interval by examining 100fs of the circumferential (Item No. Bl.11) and 10085 of the-1ongitudinal (Item No. Bl.12) shell welds in the reactor vessel. These augmented examinations may also be performed during the last period of the second inspection interval." Paragraph 2.11.1 " Reactor Vessel, Category B-A," is being deleted. Revised pages of both the Units 2_and 3 ISI programs are enclosed. The augmented examination in 50.55a(g)(6)(ii)(A) was a one time examination applicable to the ISI inspection interval in effect on September 8, 1992. On that date San Onofre Units 2 and 3 were in their first inspection interval, e V' 9407060157 940701 PDR ADOCK 05000361 l-O PDR
Document Control Desk and satisfied this rule as discussed below. Inspections for the second interval, which began on April 1, 1994, will be performed in accordance with the 1989 Edition of the ASME Code. The auamented exaaination is not applicable to the second interval. FIRST TEN YEAR INTERVAL EXAMINATIONS Examinations of the reactor vessel were performed in accordance with 50 55a(g)(4) during the Cycle 7 refueling outages at each Unit (completed in August of 1993, on U n Onofre Unit 2 and December of 1993, on San Onofre Unit 3.) These were the final outages during the first 10-year 151 interval and were examined in accordance with the 1977 Edition,1979 Addenda, of the Code. Because San Onofre Units 2 and 3 were in the first inspection interval on September 8, 1992, the requirements of the rule regarding " augmented examinations of reactor vessel" were met in accordance with paragraph 50.55a(g)(6)tii)(A)(4). In accordance with this paragraph, the examinations were performed in accordance with 50.55a(g)(4). It is our interpretation of this rule that the definition of " essentially 100%" included in 50.55a(g)(6)(ii)(A)(2) did not apply to our first inspection interval inspections. This interpretation is supported by the federal Register Notice which accompanied the rule (Vol. 57 1152,datedAugust6,1992,page34670) which states: "Section 50.55a(g)(6)(ii)(A)(4) specifies that a licensee that has either completed or scheduled an inspection of essentially 100 percent of the length of all Examination Category B-A shell welds during the inservice inspection interval in effect when the rule becomes effective does not have to implement the required augmented examination of the reactor vessel shell welds, primarily, this paragraph is intended to permit licensees who are in the first inspection interval who use the essentially 100 percent reactor vessel shell welds examination required for that interval by section XI to satisfy the requirements for the augmented examination of the reactor vessel. The technical objective of the augmented examination will be accomplished under these conditions. These licensees will continue to apply the current requirements of 650.55a(g)(4) until the next inspection interval when future examinations will be performed based on ASME section XI, 1989 Edition, or later Code edition and addenda specified in 650.55a(b)." Section 50.55a(g)(4) allows the examiner to complete the required examinations to the " extent practical within the limitations of design, geometry and materials of construction of the components." Since we satisfied this requirement, we believe we are in compliance with the rule and the definition of " essentially 100%" included in paragraph 50.55a(g)(6)(ii)(A)(2) did not apply to this inspection interval. As stated in our October letter we examined all of the reactor vessel welds; l
e Document Control Desk however, as discussed below we were not able to examine 100% of each weld. During the reactor vessel exauinations, we used a remote 151 tool to perform UT examinations from the inside surface. No outside vessel surface examinations were performed, as we determined them not to be necessary to properly comply with the inspection requirements, and thus could not justify the radiation exposure and expense associated with such an examination. All of the accessible portions of all of the shell weldt, were examined. The San Onofre Units 2 and 3 reactor vessel shells each contein 12 welds; 3 circumferential (Bl.11) and 9 longitudinal (Bl.12). We were able to examine greater than 90% of the total weld volume of all the welds but not 90% of each individual weld. This was due to design interferences presented by the flow skirt, support lugs, and specimen holders. During the course of examination we looked at each weld several times, in different angles and orientation. We did 100% of the accessible portions of each weld. On an individual weld basis, for Unit 2, we examined the three circumferential weld volumes at 85, 87, and 100 %. For the 9 longitudinal welds, two fell below the 90% level at 77 and 84 %. The remaining 7 longitudinal welds were 100% accessible. The Unit 3 welds were similarly inspected. The coverage of the three circumferential weld volumes was 96, 84, and 100%, and the coverage of the two longitudinal welds was a5 and 76%. The remaining welds were 100% accessible. Very Truly Yours, dbh Y;!' Enclosures cc: L. J. Callan, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident inspector, San Onofre Units 2 & 3 H. B. Fields, NRC Project Manager, San Onofre Units 2 and 3 _~
F AN JNol'Rli NUCLIIAR Gl NIIRATING STATION INSI.RVICII INSI'IrTION l'ROGRAM UNIT 2 PAGI! 22 OF 41 j High Pressure Safety injection System (recirculation ponion), Ixw Pressure Tafc.y injection System (shutdown cooling ponion), Reactor Coolant Sampling System (post-accident sampling piping), and Containment Spray System. .2 The VT 2 visual examination reports will cover the pressure retaining boundary up to the containment pene:.ations.
- P 2.11 Augmented Inservice Inspection Pursuant to the requirements of 10 CFR 50.55a (g)(6)(ii), augmented inservice inspection may be performed on cenain components to provide added assurance of structuml reliability. The following section dennes the component augmented inservice examination requirements for the second inspection interval.
t 2-14rl-Reaetor-Vesseh-Gategory-B-Ai The-eug mental-exa minat ion-ratuirement s-in-10-G FR40.-55a(g)(6)(ii)(A)(2)h 1993-Revisiotrof4 he-Gale 4*f-Fetieral-Regulation 9-Title-40r-Pa n40-were I. satisGal-du ring-ihe-last-periml4*f-t heferst-in spectio n-ir.tervaHiy-examining Rev i 4004efa he-eircu m ferential-{ltenrNo-B irl4)-e nd-100&ofa he-longitudinal-(Item-NorBlvl 2Hhell-welds-ina he-reactor-vessel---These-amg mented-examitmtions-may-also 4*e-perfonnedsluring-t he-lastterimis*faleseeomi-inspeetion-interval.- 2 2.I1.2 Iligh Energy Lines: .1 "No break zones" in the hiain Steam Lines and ponions of other high energy piping which penetrate containment will receive 100% ultrasonic examination of all circumferential and longitudinal welds in the area between the Orst pipe whip restmint beyond inboard and outboard containment isolation valves. 4 Piping which operates above 200'F or 275 psig is considered high energy piping. .2 The ISI Zone Drawings included in Appendix 14 define the augmented ISI boundaries for piping in the hiain Steam, hiain Feedwater, Auxiliary Feedwater, and Steam Generator Blowdown systems. .3 In addition, certain welds at postulated breakpoints on the hiain Steam line inside containment are subject to augmented ISI to provide an additional safety margin, such that elfects of jet impingement do not require further protective measures as explained in F3AR Section 3.6A.2.4.3.
._-= g SAN ONOFRE NUCl. EAR GENERATING STATION INSERYlCE INSPECTION PROGRAM l UNIT 3 PAGE 22 OF 41 Iligh Picssure Safety Injection System (recirculation ponion), Low Pressure Safety Injection System (shutdown cooling ponion), Reactor Coolant Sampling System (post-accident sampling piping), and Containment Spray System, .2 The VT-2 visual examination repons will cover the pressure retaining tmundary up to the containment penetrations. 2.11 Augmented Inservice Inspection Pursuant to the requirements of 10 CFR 50.55a (g)(6)(ii), augmented inservice inspection may be perfonned on cenain components to provide added assurance c' structural reliability. The following section dennes the component augmented a.;rvice examination requirements for the second inspection interval. 24M-Renetor-Vesselv-Gategory-B-A+ Thsmgmented-examination-wtuirements-int-14GFR 50-Sha(g)(6)(ii)(A)(2}, W9FRevisionef-the-Gtxle-of-Federal-Regulati<ms-Title-lOrhtt-50-were satisGed-during4he4ast-peritxi-of-the-first-inspection-interwd-by-examining Rev~ LOO 4-ef-the-eircumferential-(Itent4h-BM4Hnd-10mf-the4<mgitudinal-(item-No-BM 2Hhell-weldWnahe-reaetor-vessel. The<,e-augmented-examinations-may-also4)e-perfonned-during-the4ast-peri <xi+f-timend-inspeetkm4ntetvah 2.11.2 Iligh Energy Lines: .1 "No break zones" in the hiain Steam Lines and portions af other high energy piping which penetrate containment will receive 100% ultmsonic examination of all circumferential and longitudinal welds in the area between the first pipe whip restmint beyond intmard and outboard containment isolation valves. Piping which operates above 200*F or 275 psig is considered high energy piping. .2 The ISI Zone Drawings included in Appendix 14 define the augmented ISI boundaries for piping in the hiain Steam, hiain Feedwater, Auxiliary Feedwater, and Steam Generator Blowdown systems. .3 In addition, certain welds at postulated breakpoints on the hiain Steam line inside containment are subject to augmented ISI to provide an additional safety margin, such that effects of jet impingement do not require funher protective measures as explained in FSAR Section 3.6A.2.4.3. __}}