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

{{Adams
| number = ML022590091
| issue date = 09/04/2002
| title = Corrected Page 109 to Relief Request 10 Requesting Alternative to Requirement to Examine All Three Regenerative Heat Exchanger Vessels
| author name = Webb T
| author affiliation = Nuclear Management Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000266, 05000301
| license number = DPR-024, DPR-027
| contact person = 
| case reference number = NRC 2002-0077
| document type = Inservice/Preservice Inspection and Test Report, Letter type:NRC
| page count = 13
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter: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
                                                                                                       -   -- q

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}}