NRC 2002-0077, Corrected Page 109 to Relief Request 10 Requesting Alternative to Requirement to Examine All Three Regenerative Heat Exchanger Vessels

Corrected Page 109 to Relief Request 10 Requesting Alternative to Requirement to Examine All Three Regenerative Heat Exchanger Vessels

ML022590091
Person / Time
Site:	Point Beach
Issue date:	09/04/2002
From:	Webb T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2002-0077
Download: ML022590091 (13)

Text

NMC Committed to NuclearExcellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC NRC 2002-0077 10 CFR 50.55a September 4, 2002 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Ladies/Gentlemen:

Dockets 50-266 and 50-301 Point Beach Nuclear Plant, Units 1 and 2 Fourth Interval Inservice Inspection Program Relief Request No. 10 Corrected Page

Reference:

1. Letter from T. J. Webb (NMC) to Document Control Desk dated March 22, 2002, PBNP Fourth Interval Inservice Inspection Program, Plan and Schedule.

In accordance with IWA-1400(c) of the 98A00 Section XI ASME Code, Nuclear Management Company, LLC (NMC), licensee for the Point Beach Nuclear Plant (PBNP), submitted the PBNP Fourth Interval Inservice Inspection (ISI) Program in Reference 1.

Included in Appendix D of Reference 1 were several relief requests. Relief Request No. 10 requested an alternative to the requirement to examine all three regenerative heat exchanger vessels.

During a conference call on August 22, 2002, NRC identified an error in Relief Request No. 10 on page 109 of 115. In accordance with the 1998 Edition of ASME Section XI with Addenda through 2000, the requirement for Class 2 nozzle to shell welds, examination category C-B, item no. C2.21, is to perform both surface and volumetric examinations on each weld. In Relief Request No. 10, NMC summarized this requirement as only a volumetric exam. The intent was to meet the requirement of the ASME code for both surface and volumetric examinations.

In response, attached is the corrected page 109 for Relief Request No. 10.

Sincerely, Thomas J. Webb Regulatory Affairs Manager 6590 Nuclear Road 0 Two Rivers, Wisconsin 54241 Telephone: 920.755.2321

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NRC 2002-0077 Page 2 LAS/kmd Attachment cc: NRR Project Manager NRC Resident Inspector NRC Regional Administrator PSCW

Attachment To the Letter From:

T. J. Webb (NMC)

To:

Document Control Desk (NRC)

NRC 2002-0077 Fourth Interval Inservice Inspection Program Relief Request No. 10 Corrected Page 109 Point Beach Nuclear Plant

POINT BEACH NUCLEAR PLANT ISI CL 1,2,3 PROGRAM INSERVICE INSPECTION PROGRAM Revision 0 March 11, 2002 PBNP CLASS 1, 2, AND 3 INSERVICE INSPECTION PROGRAM RHE-N8-IRS RHE-N8-IRS RHE-N9-IRS RHE-N9-IRS RHE-N12-IRS RHE-N12-IRS Examination Category C-A, Item No. C1.20, tubesheet to shell weld, volumetric examination Unit 1 Unit 2 RHE-04 RHE-04 RHE-08 RHE-08 RHE-12 RHE-12 Examination Category C-A, Item No. C 1.30, tubesheet to shell weld, volumetric examination Unit 1 Unit 2 RHE-03 RHE-03 RHE-07 RHE-07 RHE-1I RUE-i1 Examination Category C-B, Item No. C2.21, tubesheet to shell weld, surface and volumetric examination Unit 1 Unit 2 RHE-N2 RHE-N2 RHE-N3 RHE-N3 RHE-N6 RHE-N6 RHE-N7 RHE-N7 RHE-NI1 RUE-NI1 RHE-N11 RUE-N11 Relief Requested Relief is requested from performing the examinations of the Regenerative Heat Exchanger welds as required by the 1998 Edition of Section XI with Addenda through 2000.

Basis for Relief The Regenerative Heat Exchanger is a high radiation component, located inside of a lock high radiation area. It is the greatest single source of radiation exposure accumulated during a normal refueling outage for ISI and support personnel. Just as an outage begins, Radiation Protection personnel make a survey of the area to document dose rates. These rates are typically 700 mr to 1400 mr for the general area.

Hot spots of 3000 mr are normally found on contact with the heat exchanger. The following dose accumulations are expected using 3.0 Rem-hour due to the close contact the workers and NDE examination personnel experience in the course of performing their duties for each weld:

0.2 Man-hours for insulation removal = 0.6 Man-Rem 0.2 Man-hours for weld cleaning and preparation = 0.6 Man-Rem 0.75 Man-hours for conducting examinations = 1.5 Man-Rem

NMC Committed to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC NRC 2002-0074 September 4, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Ladies/Gentlemen:

Docket 50-266 Point Beach Nuclear Plant, Unit 1 Summary of the Spring 2001 Unit 1 (U1R26) Steam Generator Eddy Current Examinations In accordance with the requirements of Point Beach Nuclear Plant Technical Specifications 5.6.8.b.1 and 5.6.8.b.2, Nuclear Management Company, LLC (NMC), licensee for PBNP, is submitting the summary of the spring 2001 Unit 1 Steam Generator Eddy Current Examinations.

The attached summary reports the number and extent of tubes tested and the location and percentage of wall-thickness penetration for each indication of degradation in each steam generator. Note that these descriptions were previously provided in the Annual Results and Data Report for PBNP that was required by the previous Technical Specifications. The annual report for the year 2000 was submitted on February 28, 2001. With the implementation of Improved Technical Specifications at Point Beach in 2001, this annual report is no longer required.

Please contact us if you have any questions.

Sincerely, Thomas J. Webb Regulatory Affairs Manager LAS/kmd Attachment 6590 Nuclear Road 0 Two Rivers, Wisconsin 54241 Telephone- 920 755.2321

N4RC 2002-0074 Page 2 cc: NRR Project Manager NRC Resident Inspector NRC Regional Administrator PSCW

Attachment 1 To the Letter From:

T. J. Webb (NMC)

To:

Document Control Desk (NRC)

NRC 2002-0074 Summary of the Spring 2001 Unit I (U1R26)

Steam Generator Eddy Current Examinations Point Beach Nuclear Plant Unit I

NRC 2002-0074 Page 1 of 6

SUMMARY

OF THE SPRING 2001 UNIT 1 (U1R26) STEAM GENERATOR EDDY CURRENT EXAMINATIONS During the Point Beach Nuclear Power Plant Unit 1 spring 2001 refueling outage (U1R26), the following steam generator (SG) services were performed:

Eddy Current Examinations The U1R26 SG tube eddy current examination program included:

1) A bobbin coil examination of 100% of the non-plugged tubes through their entire length.

2) A low frequency rotating plus point examination of 20% of the row 1 U-bends, from the sixth hot leg tube support plate to the sixth cold leg tube support plate.

3) A high frequency rotating plus point examination of 20% of the row 1 U-bends, from the sixth hot leg tube support plate to the sixth cold leg tube support plate.

4) A rotating plus point examination of 40% of the hot leg tube expansion transitions, 2 inches above and below the secondary face of the tubesheet.

5) Supplemental rotating plus point examinations of ambiguous bobbin coil signals, as required.

As required by Technical Specification 5.6.8.b.1, Table 1 contains a summary of the U1R26 SG eddy current examinations, including the number and extent of tubes tested.

As required by Technical Specification 5.6.8.b.2, Table 2 contains the location and percent of wall-thickness penetration for each indication of degradation in SG A, and Table 3 contains the location and percent of wall-thickness penetration for each indication of degradation in SG B.

Steam Generator Repairs SG repairs were performed during the U1R26, including tube plug removal and replacement and SG tube plugging.

As required by Technical Specification 5.6.8.b.3, Table 4 lists the tube locations that were plugged or repaired.

NRC 2002-0074 Attachment 1 Page 2 of 6 TABLE 1

SUMMARY

OF THE U1R26 STEAMU GENERATOR EDDY CURRENT EXAMINATIONS Item Scope Extent Number of Tubes Tested SGA SGB 1 Bobbin Examination TEH to TEC 3210 3209 2 Low Row U-Bends (Low Freq) 06H-1" to 06C-1" 19 19 3 Low Row U-Bends (High Freq) 06H-l" to 06C-1" 17 19 4 Hot Leg Tubesheet Expansion Transitions TSH +/- 2" 1308 1300 5 Supplemental RPC Testing TSH-1" to FBH+I" 4 0 6 Supplemental RPC Testing TSH-1" to 01H+1" 0 3 7 Supplemental RPC Testing FBH-l" to 01H+1" 0 1 8 Supplemental RPC Testing 01H-1" to 02H+1" 2 2 9 Supplemental RPC Testing 02H-1" to 03H+l" 2 0 10 Supplemental RPC Testing 03H-1" to 04H+1" 1 0 11 Supplemental RPC Testing 04H-l" to 05H+1" 1 1 12 Supplemental RPC Testing 05H-1" to 06H+l" 2 2 13 Supplemental RPC Testing 06H-l" to 06H+l" 0 1 14 Supplemental RPC Testing 06C-1" to 06C+1" 1 1 15 Supplemental RPC Testing 06C+1" to 05C-1" 2 4 16 Supplemental RPC Testing 05C+1" to 04C-1" 2 0 17 Supplemental RPC Testing 04C+1" to 03C-1" 3 0 18 Supplemental RPC Testing 03C+1" to 02C-1" 1 2 19 Supplemental RPC Testing 02C+1" to O1C-I" 1 3 20 Supplemental RPC Testing 01C+I" to FBC-1" 1 0 21 Supplemental RPC Testing FBC+1" to TSC-1" 0 4 22 Supplemental RPC Testing 01C+1" to TSC-I" 1 0

NRC 2002-0074 Page 3 of 6 Table 1 Nomenclature FBC: Flow distribution baffle cold leg side FBH: Flow distribution baffle hot leg side RPC: Rotating Pancake Coil (Plus Point)

TEC: Tube end cold leg side TEH: Tube end hot leg side TSC: Top of tubesheet cold leg side TSH: Top of tubesheet hot leg side O1H: First hot leg tube support plate 02H: Second hot leg tube support plate 03H: Third hot leg tube support plate 04H: Fourth hot leg tube support plate 05H: Fifth hot leg tube support plate 06H: Sixth hot leg tube support plate 06C: Sixth cold leg tube support plate 05C: Fifth cold leg tube support plate 04C: Fourth cold leg tube support plate 03C: Third cold leg tube support plate 02C: Second cold leg tube support plate O1C: First cold leg tube support plate

NRC 2002-0074 Page 4 of 6 TABLE 2 LOCATION AND PERCENT OF THICKNESS PENETRATION U1R26 STEAM GENERATOR A ROW COL  % LOCATION ROW - -COL - % - 'LOCATION 32 14 3 AV3 45 43 12 AV1 40 27 4 AV3 34 65 12 AV4 40 25 5 AV2 27 71 12 AV3 31 63 5 AV3 33 37 13 AV4 33 66 5 AV3 40 47 13 AV3 38 22 6 AV3 45 49 14 AVI 45 41 6 AV1 19 54 14 AV2 45 41 6 AV4 31 63 14 AV2 34 65 6 AV3 33 66 14 AV1 27 71 6 AV4 34 69 14 AV2 34 33 7 AV2 35 56 15 AV1 34 69 7 AV1 19 54 16 AV4 39 68 8 AV4 33 71 17 AV2 32 71 9 AV2 38 54 18 AV3 34 33 10 AVI 19 61 18 AVI 40 44 10 AV3 38 43 19 AV1 33 71 10 AV3 38 43 19 AV2 33 66 11 AV2 19 61 23 AV2 27 71 11 AV2 35 56 27 AV2 Location Nomenclature AVI: Antivibration bar number 1 AV2: Antivibration bar number 2 AV3: Antivibration bar number 3 AV4: Antivibration bar number 4

NRC 2002-0074 Page 5 of 6 TABLE 3 LOCATION AND PERCENT OF THICKNESS PENETRATION U1R26 STEAM GENERATOR B ROW- COL LOCATION 23 33 6 AV1 32 44 7 AV3 22 58 8 AV1 19 36 9 AV3 32 46 11 AV2 32 49 11 AV2 32 32 12 AV3 22 58 12 AV4 23 33 13 AV2 32 49 14 AV1 32 70 14 AV1 45 44 14 AVI 32 46 15 AV3 22 58 16 AV3 33 71 17 AV1 32 70 17 AV2 23 33 17 AV3 45 46 19 AV1 22 58 19 AV2 32 38 20 AV1 32 38 27 AV4 32 38 30 AV2 32 38 39 AV3 Location Nomenclature AV1: Antivibration bar number 1 AV2: Antivibration bar number 2 AV3: Antivibration bar number 3 AV4: Antivibration bar number 4

NRC 2002-0074 Page 6 of 6 TABLE 4 IDENTIFICATION OF TUBES PLUGGED OR REPAIRED U1R26

-SG ROW COL PCT LOCATION P/R B 1 1 NA NA R B 2 1 NA NA R B 43 40 NA NA R B 32 38 39 AV3 P Nomenclature AV3: Antivibration bar number 3 P: Plugged preventatively during U1R26 R: Repaired during U1R26. Repair consisted of removal of old Westinghouse plug and PIP assembly and replacement with an Inconel 690 rolled plug