View source for JPN-98-039, Forwards Response to Request for Addl Info Re Response to GL 92-01, Reactor Pressure Vessel Integrity. Ltr Includes New Commitments

{{Adams
| number = ML20151S244
| issue date = 08/31/1998
| title = Forwards Response to Request for Addl Info Re Response to GL 92-01, Reactor Pressure Vessel Integrity. Ltr Includes New Commitments
| author name = Knubel J
| author affiliation = POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000333
| license number = 
| contact person = 
| document report number = GL-92-01, GL-92-1, JPN-98-039, JPN-98-39, TAC-MA1190, NUDOCS 9809040040
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 5
| project = TAC:MA1190
| stage = Request
}}

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123 Main Street White Pla ns, New York 1c301 914 681.6950 914 287.3300 (Fax)
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August 31,1998                               i JPN-98-039 United States Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555                                                                                   I

==Subject:==
James A. FitzPatrick Nuclear Power Plant                                              l Docket No. 50-333 Request for Additional Information Regarding Response to GL 92-01 Reactor Pressure Vessel Intearity (TAC No. MA1190)

==References:==
: 1.         NRC letter, J. F. Williams to J. Knubel, dated June 1,.1998                ;
regarding request for additional information regarding reactor             l pressure vessel integrity at the James A. FitzPatrick Nuclear Power Plant (TAC No. MA1190).                                              )
: 2.         NYPA letter, J. Knubel to USNRC, (JPN-98-008), dated March 9,1998 regarding " Revised Reactor Vessel Material Surveillance Program Summary Report and Implementation Schedule."
: 3.         NYPA letter, J. Knubel to USNRC, (JPN-97-035), dated                   j November 10,1997 regarding " Reactor Pressure Vessel Material Surveillance Program Summary Report and                         I implementation Schedule."

==Dear Sir:==

The Authority's response to the NRC staff's request for additionalinformation (Reference 1) regarding reactor pressure vessel integrity is attached.
Our review shows that the limiting reactor vessel material for FitzPatrick has not i
changed, nor did the reference temperature increase. No chemistry data other than that previously considered by the Authority were identified. The limiting weld has not changed and the best-estimate copper and nickel values remain unchanged from those reported to the staff in 1998 (Reference 2). Consequently, no changes to the prusure-temperature (P-T) limits are required as a result of changes in material chemistry, 8
i       9909040040 900831 PDR     ADOCK 05000333 l       P                           PDR      ,

    .                                                                                    i l
1 1
As stated in the 1997 reactor vessel material surveillance program summary report (Reference 3), the Authority will prepare and submit a proposed change to the FitzPatrick Technical Specification to reflect new P-T limits not later than June 1999. This change will extend the P-T curves from the current limit of 16 EFPYs l
(effective full power years) to 24 and 32 EFPYs to reflect the effects of reactor  i fluence and the latest test capsule results.
j This letter includes no new commitments if you have any questions, please contact Ms. Charlene Faison.
Very truly y rs, f ~
v                     .
1 Jim Knubel               l Sr. Vice President and Chief Nuclear Officer cc:
Regional Administrator United States Nuclear Regulatory Commission                                        !
475 Allendale Road King of Prussia, PA 19406                                                          l Office of the Resident inspector                                                   !
United States Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Joseph Williams, Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 United States Nuclear Regulatory Commission Mail Stop 1482 Washington, D.C. 20555 Attachments:
: 1. James A. FitzPatrick Nuclear Power Plant, Response to NRC June 1,1998 Request for Additional Information Regarding Reactor Pressure Vessel Integrity i
2

L*                                   Attachmsnt I to JPN-98-039 l                             James A. FitzPatrick Nuclear Power Plant                                -
.                   Response to RAI Regarding Reactor Pressure Vessel Integrity                       ;
1              .
Introduction
?                                                                                                      l The Authority's response to the NRC staff's June 1,1998 request for additional information regarding reactor pressure vessel integrity (Reference 1) is below.              .
Our review shows that the limiting reactor vessel material for FitzPatrick has not changed, nor did the reference temperature increase. No chemistry data other than that previously considered by the Authority were identified. The limiting weld has not changed and the best-estimate copper and nickel values remain unchanged from those reported to the staff in 1998 (Reference 2).
l_       Most of the information requested is contained in references 2,4,7 and 8.
Reauest 1 An evaluation of the bounding assessment in ine reference above (Undate of Boundina Assessment of BWR/2-6 Reactor Pressure Vessel Intearity issues, BWRVIP-46. December 1997) and its applicability to the determination of the best-estimate chemistry for all of your RPV beltline welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Table 1 for each RPV beltline weld material, if the limiting material for your vessel's P-T limits evaluation is not a weld, include the information requested in Table 1 for the limiting material also.
Response 1 in a {{letter dated|date=March 9, 1998|text=March 9,1998 letter}} to the NRC staff (Reference 2), the Authority documented             ;
its evaluation of the BWRVIP-46 report (Reference 4). The Amhmity's evaluation included a review of any "new" data contained in the report - e.g. data not                   ,
previously reported. As stated in the March 1998 letter, the data contained in                I BWRVIP-46 does not affect the current pressure-temperature (P-T) curves for boiling water reactors (including FitzPatrick) due to weld chemistry variability.
A General Electric report (Reference 5) summarizing the results of tests conducted on the latest reactor vessel material surveillance capsule (capsule no. 2) was submitted to the NRC in 1997 (Reference 3). The GE report was subsequently
        - updated (Reference 7) and resubmitted to the NRC staff (Reference 2) in 1998. In these reports, beltline weld material chemistry values were revised from the chemistry values submitted in 1996 (Reference 6) to reflect the best estimate values in the 1997 Combustion Engineering Owners Group report, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," (Reference 8).
The weld material chemistry values appearing in Table 7-1 of the 1998 surveillance capsule 2 report (Reference 2), were taken from the CEOG report (Reference 8).
The CEOG report (Reference 8) describes the methodology used to arrive at the best estimate values.
I 1
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                                        ' Attachment I to JPN-98-039 James A. FitzPatrick Nuclear Power Plant Response to RAI Regarding Reactor Pressure Vessel Integrity The 1998 updated surveillance capsule report (Reference 2) shows that, for the FitzPatrick plant, the limiting material continues to be axial weld 2-233, heat 27204/12008.
Table A compares th.e 1996 (Reference 6) and 1998 (Reference 2) copper and nickel values.
Table A Comparison of Cooper /Nicket Content For Selected Weld Metal Heats as Reoorted in the 1996 and 1998 Submittats Heat N*>.                     Cu/Ni Content       Cu/Ni Content       i 1996 Response to       1998 Surveillance Rev.1, Supp.1 to          Capsule 2 Test      '
GL 92-01              Report          i 13253/12008               Cu             0.253 %              0.210 %          l Ni              O.804 %             0.873 %          !
27204/12008*               Cu             0.183 %              0.219 %          !
Ni              1.008 %             0.996 %          I 305414                Cu             0.337 %              0.337 %
Ni             0.600 %              0.609 %
* Limiting material is in axial weld 2-233.
The latest beltline weld material chemistry values and the additional weld data requested by the NRC staff (Reference 1, Table 1) is provided in Table B.
l l-                                                           2

e Attcchment I to JPN-98-039                                                                                                                                                                         ,
James A. FitzPatrick Nuclear Power Plant Response to RAI Regarding Reactor Pressure Vessel Integrity Table B -Information Requested on RPV Weld and/or Limiting Materials                                                                                                                                -
Facility: James A. FitzPatrick Nuclear Power Plant Vessel Manufacturer: Combustion Engineering RPV            Best-                                                 Best-        EOLID        Assigned                           Method of                                                                                                      Initial  6,   6A  Margin  ART or W eld                 Estimate Estimate                                                  Fluence        Materia!                    Determining                                                                                                          RTuor                    U nor at Wire               Copper                                                 Nickel      (x 10")       Chemistry                                              CFm                                                                                (RTworu)                         EOL (See Heat "'                                                                                                 Factor (CF)                                                                                                                                                                  Note A)_
27204/                   O.219                                                 0.996        0.181          231.0                                RG 1.99,                                                                                                   -48"F   O.0 28.0   56      109 12008                                                                                                                              Rev.2, Tbl.1 13253/                   0.210                                                 0.873        0.181          208.7                                RG 1.99,                                                                                                   -50 F   O.0 .28.0  56      105 12008                                                                                                                              Rev.2, Tbl.1 305414                    0.337                                                 0.609        0.181          209.1                               RG 1.99,                                                                                                    -50 F   O.0 28.0   56      100 Rev.2, Tbl.1 (1) or the material identification of the limiting material as requested in Section 1.O(1.)
(2) deterrnined from tables or from surveillance data Discussion of the Analysis Method and Data Used for Each Weld Wire Heat Weld Wire Heat                                                                                    Discussion All                                                                                               CE NPSD-1039, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CEOG Task 902, June 1997.
Note A: No unirradiated data was available. Predictions of Regulatory Guide 1.99, Rev. 2 position 1, were used to calculate ART.
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Attachment I to JPN-98-039 James A. FitzPatrick Nuclear Power Plant Response to RAI Regar'd ing Reactor Pressure Vessel Integrity Reauest 2 i       If the limiting material for your plant changes or if the adjusted reference
).      temperature for the limiting materialincreases as a result of the above evaluations,
: l.      provide the revised RTu,value for the limiting material. In addition, if the adjusted
;       RTu, value increased, provide a schedule for revising the P-T limits. The schedule l       should ensure that compliance with 10 CFR Part 50 Appendix G is maintained.
Response 2 '
L i
L       The limiting reactor vessel material for FitzPatrick has not changed, nor did the                  !
L       adjusted reference temperature increase. No chemistry data other than that l
previously considered by the Authority were identified. The limiting weld has not changed and the best-estimate copper and nickel values remain unchanged from those reported to the staff in 1998 (Reference 2). Consequently, no changes to the l       pressure-temperature (P-T) limits are required as a result of changes in material                   ;
chemistry.                                                                                          !
in the 1997 program summary and implementation schedule (Reference 3), the l       Authority committed to prepare and submit a proposed change to the FitzPatrick l
Technical Specification to reflect new P-T limits not later than June 1999.' This
                                                                                                           )
schedule complies with the requirements of 10 CFR Part 50, Appendix G.
The P-T limit curves in the current FitzPatrick Technical Specifications (Figures 3.6-1,2 and 3) are valid to 12,14 and 16 effective full-power years (EFPYs). These curves were proposed by the Authority (Reference 9) and subsequently approved by the NRC as amendment 158 to the FitzPatrick Technical Specifications (Reference                     ;
10).
A comparison of the P-T curves currently in the FitzPatrick Technical Specifications and those calculated based on the 1998 capsule test report show that the existing l       curves are conservative and over-estimate the effects of radiation on reactor l       materials. In fact, the new 24 EFPYs curves are less limiting than those currently in the Technical Specifications, which are only valid to 16 EFPYs. (Note that the P-T curves in the 1998 capsule test report have not been subjected to final acceptance review by the Authority and consequently may change.)
  ;                                                 4

Attachment I to JPN-98-039 James A. FitzPatrick Nuclear Power Plant Response to RAI Regarding Reactor Pressure Vessel Integrity References
: 1. NRC letter, J. F. Williams to J. Knubel, dated June 1,1998 regarding " Request for AdditionalInformation Regarding Reactor Pressure VesselIntegrity at the James / FitzPatrick Nuclear Power Plant (TAC No. MA1190.)"
                              ~
: 2. NYPA letter, J. Knubel to USNRC, (JPN-98-008), dated March 9,1998 regarding
          " Revised Reactor Pressure Vessel Material Surveillance Program Summary Report and implementation Schedule."
: 3. NYPA letter, R. J. Deasy to USNRC, (JPN-97-035), dated November 10,1997 regarding " Reactor Pressure Vessel Material Surveillance Program Summary Report and implementation Schedule."
: 4. Update of Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity issues, (BWRVIP-46), EPRI 109727, December 1997.
: 5. General Electric Report GE-NE-B1100732-01, Rev.0, " Plant Fitzpatrick RPV Surveillance Materia!s Testing and Analysis of 120 CapsuM at 13.4 EFPY,"
October 1997.                                                          .
: 6. NYPA letter, W. J. Cahill, Jr. to USNRC, (JPN-96-020), dated May 2,1996 regarding " Generic Letter 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity - Revised Six Month Response."
: 7. General Electric Report GE-NE-B1100732-01, Rev.1, " Plant Fitzpatrick RPV Surveillance Materials Testing and Analysis of 120 Capsule at 13.4 EFPY,"
February 1998.
: 8. CE NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CEOG Task 902, June 1997.
: 9. NYPA {{letter dated|date=January 12, 1990|text=letter dated January 12,1990}} (JPN-90-007) regarding proposed changes to the FitzPatrick Technical Specifications - Pressure Temperature Limits (JPTS-89-012).
10 NRC letter, D. E. LaBarge to J. C. Brons dated April 26,1990 regarding issuance of Amendment No.158 to FitzPatrick Technical Specifications (TAC No. 75870).
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