ML20150A712

Forwards Addl Info Re Inservice Insp Program,Per 880119 Telcon.Attachment Contains Info on Util Commitment to Volumetrically Examine 7.5% Sample of Welds in Containment Spray Sys & Revised Relief Request Re HXs

ML20150A712
Person / Time
Site:	Byron
Issue date:	03/09/1988
From:	Ainger K COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8803160017
Download: ML20150A712 (8)

Text

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's Commonwealth Edison

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• One Firtt National Plaze., Chicago, Illinois v- i.

v Address Reply to: Post Offc3 Box 767

(, Chicago,irrinois 60690 0767 March 9, 1988 U.S. Nuclear Regulatory Commission Attn: Document Control Desk washington, DC. 20555

Subject:

Byron Station Units 1 and 2 Inservice Inspection Program NRC Docket Nos. 50-454 and 50-455

Reference:

(a) May 28, 1987 letter from K. A. Ainger to U.S. NRC Gentlemen:

A teleconference was held on January 19, 1988 between Commonwealth Edison personnel and the NRC staff to discuss the Byron Inservice Inspection (ISI) Program. As a result of this teleconference, Commonwealth Edison agreed to provide the NRC staff additional information regarding the ISI Program. Attachment A contains the following information:

1) A commitment to volumetrical?v examine a 'i.5% sample of welds in the containment spray system.

2) A revised relief request NR-12 regarding the residual heat removal heat exchangers.

3) A description of the changes made to the program beyond those described in reference (a).

4) A description of revised relief request NR-3 regarding the pressurizer lower nozzle and steam generator primary nozzles.

Please direct any further questions regarding this matter to this office.

Very truly yours, K. A. Ainger Nuclear Licensing Administrator

/klj att.

cc: Byron Resident Inspector NRC Region III Office 4337K 8803160017 080309 Yy ADOCK 05000454 ^0 PDR q PDR fl j t t

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ATTACHMENT A

1) Byron Unit 2 components have been incorporated into the ISI Program.

2) The letdown reheat heat exchanger, ICV-05-A, weld LRHXC-01 was put into the Unit-1 table under Code Category C-H, Item number C1.20. This weld was inadvertently omitted from the previous revision.

3) The regenerative heat exchangers, ICV 03AA and ICV 03AB, welds RGXN-01 and RGXN-02, were moved from Code Category C-B, Item number C2.ll to Item number C2.21. The heat exchangers have a nominal wall thickness of greater than one-half inch.

4) In Code Category C-H, Item numbers C7.10, C7.20, C7.30, C7.40, C7.50, C7.60, C7.70, and C7.80, leak testing exemptions were taken on specific components pursuant to ASME Section XI, Table IWC-2500-1, note 7.

5) In Code Category D-A, Item number Dl.10, the boric acid processing system, the boron thermal regeneration system, the chemical and volume control system, the process sampling system, the reactor building equipment drain system, and the auxiliary building equipment drain system were taken out of this Code category because the Class 3 Portions of these systems do not support reactor shutdown.

6) In Code Category D-B, Item number D2.10, the chemical and volume control system,_the radioactive waste gas system, the residual heat removal system, and the safety injection system were taken out of this Code category because the Class 3 portions of these systems do not support emergency core cooling, containment heat removal or atmosphere cleanup, or residual heat removal.

7) Note 1 was incorporated into Relief Request NR-1.

8) Note 7, for turbine rotor inspections, was revised to schedule rotor inspections pursuant to Westinghouse methodology.

9) Note 9 was added to change Code Category C-F, Item number C5.11, from surface examinations to volumetric examinations for the welds selected for examination in the first interval. This was done per NRC request.

10) Note 10 was added to perform volumetric examination of a 7.5% sample selection of containment spray (CS) system welds. The examination boundaries go from one CS pump discharge to its containment isolation l penetration, for both units. This was done per NRC request. l

11) Note 11 was added because it is Unit 2 specific.

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12) _ In thn Unit 1 sx:mpt compon:nt table, instead of listing all tha Clars 2 chemical and volume control system lines as exemptions, the table was changed to state all lines are exempt. This makes the table clearer.

The same thing was done for the Class 2 portions of the reactor coolant system and pressurizer, and the Class 3 portions of the auxiliary feedwater system.

13) Relief Request NR-3 was reviewed, per NRC request. Byron Station will revise the Basis for Relief section of the Relief Request. The 5/16 inch notch currently referenced in the Basis for Relief is misleading since that particular experiment was conducted at a BWR.

The Byron Station Technical Staff and the Chief Level III NDE Examiner studied the pressurizer nozzle inner radii for ultrasonic test method-feasibility. The steam generator primary side nozzle inner radii were also studied for ultrasonic test method feasibility.

The pressurizer lower head nozzle (PN-01) is blocked from ultrasonic f test methods by the heater penetrations in the pressurizer. Therefore, this surface cannot be ultrasonically tested. Ultrasonic test methods or techniques are not available for the pressurizer upper head nozzles (PN-02, pN-03, PN-04, PN-05, and PN-06). A pressurizer upper head mock-up does not exist to allow experimentation with ultrasonic test methods and techniques. The primary concern with the nozzle inner radii is cracks forming on the inner radius of a nozzle. The VT-1 examination should identify these cracks. A remote VT-1 examination exposes personnel to less radiation than an ultrasonic test.

The steam generator primary side nozzles have a closure ring on the inner radii (N-1A and N-1B). The closure ring provides support for some types of nozzle dams used in the steam generators. This closure ring makes the ultrasonic test method difficult, if not impossible, to '

perform. The primary concern with the nozzle inner radii is cracks forming on the inner radius of a nozzle. The VT-1 examination should identify these cracks. The VT-1 examination is much more practical since someone is in the steam generators, typically, every refueling outage.

14) Relief Requests NR-4, NR-5, NR-6, and NR-7 were updated to reflect r current technology.

15) Relief Requests NR-10 and NR-11 were edited to reflect the changes made to code category C-H, Item number C7.10.

16) Relief Request NR-12 was updated to include a detailed description of the geometric constraints for examination of the residual heat removal i system heat exchanger, RH-02-AB, weld numbers RHXN-01 and RHXN-02. A copy of this Relief Request is attached for NRC review.

17) Relief Requests NR-16 and NR-17 were added because they are Unit 2 specific. ,

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l R v. 2 RELIEF Riiiwuzar NR-12 ,

1. SYSTEM: Reactor Coolant (Steam Generator, Secondary Side); Residual Heat Removal (Residual Heat Removal Heat Exchanger)

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2. NUMBER OF ITEM 5: 8 Component Number Item Number Attachment Numbers IRC-01-BA SGN-02, SGN-03 1&2 2RC-01-BA SGN-02, SGN-03 3 1RH-02-AB RHXN-01, RHXN-02 1&2 2RH-02-AB RHXN-01, RHXN-02 3

3. ASME CODE CLASS: 2

4. ASMB CODE SECTIOf XI REQUIREMENTS: Subsection INC, Table INC-2500-1, Examination Category C-B, Item C2.22 requires volumetric examination of the regions described in Figure INC-2500-4(a) or (b), for nossle inner radii in nossles without reinforcing plate in vessels > 1/2 in, nominal l thickness. Item C2.30 requires surface and volumetric examination of l the regions described in Figure INC-2500-4(c) lor nossles with Reinforcing Plate in vessels > 1/2 inch nominal thickness.

Examinations shall be conducted on nossles at terminal ends of piping l runs selected for examination under Examination Category C-F, each inspection interval. In addition, Examination Category C-H, Item C7.10 )

requires a system leakage test, (INC-5221) each inspection period for i pressure vessel pressure retaining components. l

5. BASIS FOR RELIEF: The nossles listed above contain inherent geometric (L constraints which limit the ability to perform =maningful ultrasonic examinations. The main steam nossle (SGN-03) was designed with an internal multiple venture type flow restrictor with an equivalent throat diameter of 16 in., see Attachment 1. This design is used to limit the flow in the event of a postulated steam line break. This design does not l utillse a radiused nossle as described in figures INC-2500-4(a) or j INC-2500-4(b), but instead has seven individual inner radii, corresponding to each venturi. None of which could be examined by ultrasonic examination. The main feedwater nossle (338-02) also has an internal multiple venturi type flow restrictor, and, in addition, an internal thermal sleeve, see Attachment 2. This design could not be examined due to the geometry of the nossle's internal design. The Residual Heat Removal Heat Exchanger is approximately 7/8 in, nominal wall thickness with nossles of 14 inch diameter and approximately 3/8 inch in r m inal wall thickness. Attachment 3 shows the actual i nossle-to-shell weld configuration for the Residual Heat Removal Heat

. Exchangers primary side nossles. This configuration is best characterised as a fillet welded nossle, which is most cicsely approximated by Figure INC-2500-4(c), and, thereby, is not analogous to a full penetration butt welded nossle. The examination requirement associated with this figure, with the inside of the vessel inaccessible, is a surface examination of the nossle-to-shell weld. In i addition, the inner radius of the reinforcement pad would be I representative of the nossle inner radius required for inspection. The

inherent geometric constraints of the nossle design prevent the performance of the required ultrasonic examinations of the t

, nossle-to-shell weld and the nozzle liner radius.

(0088D/0034D 2.7 - page 63 of 80 j i

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6. ALTEIOD@E TEST ME1 HOD: Visual examination (VT-1) shall be performed i either directly or remotely to the extent practical when disassembly is required for maintenance purposes not to exceed once per inspection ,

( interval. In addition, visual examination (VI-2) shall be performed )

each inspection period on all pressure retaining components.

7. JUSTIFIGTIC41 The VT-1 examination will assure early detection of detrimental flaws. Therefore, in performing the proposed alternative examinations during disassembly for maintenance, an adequate level of structural integrity can be assured for continued plant operation.

8. APPLICABLE TIME PERIOD: This relief will be required for the first 120 month inspection interval.

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(0088D/0034D 2.7 - page 64 of 80

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