View source for JPN-88-035, Advises That Util Intends to Apply Provisions of ASME Code Case N-236-1 to Mods Planned for 1988 Refueling Outage. Justification for Deferral of Leakage Testing for Reactor Water Level & Containment Spray Mods Encl

{{Adams
| number = ML20151N043
| issue date = 07/22/1988
| title = Advises That Util Intends to Apply Provisions of ASME Code Case N-236-1 to Mods Planned for 1988 Refueling Outage. Justification for Deferral of Leakage Testing for Reactor Water Level & Containment Spray Mods Encl
| author name = Brons J
| author affiliation = POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
| addressee name = 
| addressee affiliation = NRC OFFICE OF ADMINISTRATION &amp; RESOURCES MANAGEMENT (ARM)
| docket = 05000333
| license number = 
| contact person = 
| document report number = JPN-88-035, JPN-88-35, NUDOCS 8808080091
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 13
| project = 
| stage = Other
}}

=Text=
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White Piams, New Yor k 10001 914 681 6240 A NewYorkPower                                                                          aosa c.aree-E xecutive Vice President 4# Authori                                                                               ~ -e a eee - ,
1 July 22 , 1988 JPN-88-035 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk hiail Station PI-137 Washington, D.C. 20555

==SUBJECT:==

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Application of ash!E Code Case to Plant hfodifications

==References:==
: 1. ash 1E Code Case N-236-1 "Repair and Replacement of Class hic Vessels Section XI, Division 1."
: 2. NYPA letter, J. P. Bayne to T. A. Ippolito (NRC), JPN-81-95, dated Nosember 25,1981.

==Dear Sir:==

The Authority intends to apply the provisions of ash 1E Code Case N-236-1 (Reference 1) to three modifications planned for installation during the FitzPatrick 1988 refueling outa8e. Application of section 2222, "Deferral of Leakage Tests,"
allows the installation of the modifications without having to perform a leakage test as would otherwise be required by the plant Technical Specifications and 10 CFR 50 Appendix J.
l ash 1E Code Case N-236-1 states that leakage tests for welds attaching penetrations                        ,
one inch diameter or smaller may be deferred until the next scheduled leakage test.                             i For several welds associated with three modifications, the applicable leakage test is the Type A Primary Containment Integrated Leakage Rate Test (PCILRT). The Authority is not planning to perform a PCILRT during the 1988 refueling outage, it is impractical to schedule the PCILRT solely to test these welds.                     Normally, modifications of this type would receive a Type C Local Leak Rate Test (LLRT).
Ilowever, these particular welds cannot be isolated from the Primary Containment as is                          i necessary to perform an LLRT.
The next PCILRT is scheduled to be conducted during the 1990 refueling outage.
The integrity of the welds over one operating cycle is assured by visual and dye penetrant examinations conducted during installation. Delaying the leakage test by one cycle does not result in a significant reduction in any margin of safety.
The Authority previously applied this Code Case to modifications performed at FitzPatrick (Reference 2). Since that time, the Code Case was endorsed by the NRC in Regulatory Guide 1.147 and issued as part of the AShiE Code. The Authority will amend the Final Sat'ety A ialysis Report to describe the application of this Code Case 8808080091 PDR      ADOCK           h hPDC 33                                                              l P

l to the FitzPatrick plant. This will allow application of the Code Case to future plant modifications without notifying the NRC on a case by case basis.
Attachment I describes the three modifications and evaluates the applicability of the ASME Code Case to the modifications. Drawings of the modifications and affected penetrations are enclosed as Attachment 11.
Should you or your staff have any questions regarding this matter, please contact Mr. J. A. Gray, Jr. of my staff..
Very truly yours, hn   ,  rons Executive Vice President Nuclear Generation ec:      U.S. Nuclear Regulatory Commission                                                   1 475 Allendale Road King of Prussia, PA 19406 Office of the Resident Inspector U.S. Nuclear Regulatory Commission                                                   ;
P.O. Box 136                                                                         l Lycoming, NY 13093                                                                   l Mr. Harvey Abelson Project Directorate I-l Division of Reactor Projects - I/II U.S. Nuclear Regulatory Commission Mail Stop 14 B2 Washington, D.C. 20555 l
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                           -              _ _ _ _ _ _ _ _ _ _ .                                               t

NEW YORK POWER AUTIIORITY James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Attachment I to JPN-88-035 JUSTIFICATION FOR DEFERRAL OF LEAKAGE TESTING FOR REACTOR WATER LEVEL AND CONTAINhlENT SPRAY hlODIFICATIONS Three modifications are planned for installation during the 1988 refueling outage for which local leakage testing of certain welds, forming part of the primary containment pressure boundary, is impractical. The applicable leakage test for these welds is the Type A Primary Containment Integrated Leakage Rate Test (PCILRT). The Authority is not planning to perform a PCILRT during the 1988 refueling outage. It is impractical to schedule the PCILRT solely to test these welds. Visual and dye penetrant examinatione will be performed on each of the welds.
DESCRIPTION OF hlODIFICATIONS REACTOR WATER LEVEL The reacter vessel water level instrumentation system is being modified to comply with NRC Generic Letter 84-23 and Regulatory Guide !.97 Revision 2. This modification will reduce reactor water level indication errors caused by containment heat-up during postulated accident sequences and improve the reliability of the instrumentation system.
A detailed description of the modification is contained in Reference 6* and is shown in Attachment II, drawing 1.
As part of this modification, a 1 inch NPS pipe will be routed through each of existing penetrations X-26A and X-26B. These penetrations consist of 18 inch containment          '
purge piping. The instrument lines enter the purge piping inside the containment and exit the pipe outside containment, upstream of the first purge isolation valve. The Code Case is to be applied to welds located where the instrument line penetrates the 18 inch purge pipe (see Attachment II, Drawings 2 and 3). This configuration is similar to three existing containment atmosphere monitoring lines which are also routed through penetration X-26A (Reference 7).
CONTAINMENT SPR AY The containment spray system is also being modified to comply with Regulatory Guide 1.97. This modification adds instrumentation to allow control room and process            !
computer indication of containment spray flow rates. A description of the                  !
modification is contained in Reference 8 and is shown in Attachment 11, drawings 4         )
and 5.                                                                                     j i
As part of this modification, flow instruments will be installed in the 6 inch torus      l spray piping between containment penetrations X-211 A and X-211B and the inboard containment isolatic,n valves 10-MOV-38A and 10-MOV-38B, The Code Case is to be applied to the installation of the flow instrument. Attachment II, drawing 6 shows the flow instrument and its installation into the torus spray piping.
* Reference numbers refer to the list of references contained in this attachment.

I CONTAINMENT ATMOSPHERE MONITOR MODIFICATION The piping which returns the CAM samples to the torus through penetrations X-203A and X-203B are being modified as a result of removing the Beckman hydrogen and oxygen analyzers, DWA-PA and DWA-PB. FitzPatrick modification F1-87-045 entails deleting the return line and capping the connection to the torus return line. The caps are one inch or smaller in diameter and will be located on a portion of piping that cannot be isolated from the primary containment.
The existing piping configuration is shown in FSAR figures 5.2-9, sheet 3 (Reference 7). The modified piping configuration is shown in Attachment II, Drawing 7.
X-203A -      A branch line, shcwn in Attachment II, Drawing 7 as being near manual valve 201, is being cut and capped. The Code Case is applied to the cap.
X-203B -       A branch line, shown in Attachment II, Drawing 7 as being between manual valves 114 and 203, is being cut and capped. The Code Case is                    I applied to the cap.
1 A cloud is shown in Attachment 11 Drawing 7 to indicate where that Code Case is being applied.
DISCUSSION OF APPLICABILITY OF ASME CODE CASE The FitzPatrick Technical Specifications, Section 4.7 A.2.f (Reference 1), specify leakage testing requirements for containment modifications. The requirements of the j               Technical Specifications are identical to the requirements of !0 CFR 50 Appendix J,
!               Section IV.A. "Special Testing Requirements"           (Reference 2). The Technical Specifications (and Appendix J) make a distinction between major and minor modifications, but provide no guidance in defining the terms "major" and "minor."
l               The ASME Boiler & Pressure Vessel code frequently distinguishes certain repairs, l               replacements or modifications as minor by providing relaxed NDT requirements (e.g.
l Exemption from Hydrostatic Testing or Volumetric Exam). The ASME Boiler &
Pressure Vessel Code further addressed the terms "majoi" and "minor" in January, 1979, and their applicability to the leakage testing requirements of 10 CFR 50 Appendix J, via Code Case N-236 (Reference 3). This Code Case specifies that welds attaching penetrations that are 1 inch NPS or smaller are considered "minor" modifications and that leakage tests of "minor" modifications may be deferred until the l               next scheduled leikage test, provided that all other required NDE is performed. The I
Code Case was specifically intended for Class MC components, and cites the requirements of 10 CFR 50 Appendix J, Section IV.A. as supporting documentation.
The Code Case was subsequently endorsed by the NRC in Regulatory Guide 1.147, Revision 3 (Reference 4) and issued as part of the ASME Code in Section XI, IWE-5000 (Reference 5) in 1983. The case was reaffirmed and reissued as N-236-1 in September 1985 with some revisions.
The portions of containment being modified at FitzPatrick are not defined as "MC" components. However, the applicability of 10 CFR 50 Appendix J is clear. The lack of guidance in Appendix J concerning the terms "major" and "minor," combined with the fact that Type B testinh (gasketed joints) and Type C testing (isolation valves) do not apply to these modifications, makes it necessary to consult additional resources.
The Class "MC" Code Case provides a reliable source of information which I.as been l
i

endorsed by the Commission, and is therefore a valid source for determining the testing requirements for containment modifications to ANSI B31.1 components.                    l The connections in question are 1 inch NPS and therefore meet the criteria of minor             i modifications. The deferral of Type A leakage testing is justified based on these               i modifications being considered "minor" and the determination provided in Code Case              l N-236-1 and endorsed by the NRC in Regulatory Guide 1.147.
1 REFERFNCES
                                                                                                              .{
: 1)      FitzPatrick Nuclear Power Plant Technical Specifications through Amendment              i No. I14; Section 4.7.A.2.f.                                                             '
: 2)      10 CFR 50, Appendix J, Section IV.A.
: 3)      AShtE Code Case N-236 (and N-236-1).
: 4)      Regulatory Guide 1.147, Rev. 3.
: 5)      AShtE XI, IWE-5000,1983 Edition.
: 6)      Letter, J.C. Brons (NYPA) to NRC, JPN-87-033, dated June 9,              1987, concerning implementation of NRC Generic Letter 84-23,
: 7)      James A. FitzPatrick Updated Final Safety Analysis Report, Figure No. 5.2-9.
: 8)      Letter, C.A. hicNeill Jr. (NYPA) to D.B. Vassallo (NRC), JPN-84-77, dated November 30, 1984, Attachment I section 4.7, concerning inplementation of Regulatory Guide 1.97.
: 9)      Letter, C.A. hicNeill Jr. (NYPA) to D.B. Vassallo (NRC), JPN-84-77, dated November 30, 1984, Attachment I section 4.6, concerning inplementation of Regulatory Guide 1.97.
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ATTACliMENT II TO JPN-88-035 REACTOR WATER LEVEL AND CONTAINMENT SPRAY MODIFICATIONS DRAWINGS Contents i
Drawing 1 -      REACTOR WATER LEVEL INSTRUMENTATION MODIFICATIONS Drawing 2 -      REACTOR VESSEL LEVEL PENETRATION DETAIL MODIFICATION NO. F1-87-014, Sheet I of 2 Drawing 3 -      REACTOR VESSEL LEVEL PENETRATION DETAIL MODIFICATION NO. F1-8~ 314,9heet 2 of 2 Drawing 4 -     SYSTEM BLOCK DIAGRAM           -
DRYWELL/ TORUS SPRAY FLOW, MODIFICATION NO. F1-87-048, SK-87-048-1 Drawing 5 -     SYSTEM BLOCK DIAGRAM          -
DRYWELL/ TORUS SPRAY FLOW, MODIFICATION NO F1-87-048, SK-87-048-2 Drawing 6 -     ANNUBAR MODEL ANF-76 Drawing 7 -     CONTAINMENT ATMOSP11ERE MODIFICATIONS, MODIFICATION NO. F1-87-045 l
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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

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