View source for A97018, Provides Current Status of Outstanding NUREG-0737,Item II.D.1 Issues.Calculations Associated W/Unit 2 Porvs,Encl

{{Adams
| number = ML20216A870
| issue date = 08/29/1997
| title = Provides Current Status of Outstanding NUREG-0737,Item II.D.1 Issues.Calculations Associated W/Unit 2 Porvs,Encl
| author name = Mueller J
| author affiliation = COMMONWEALTH EDISON CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000295, 05000304
| license number = 
| contact person = 
| case reference number = RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM
| document report number = ZRA97018, NUDOCS 9709050153
| package number = ML20216A875
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 5
}}

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Zion Gcnerating station 10, Shiloh Ikiules.ird Zion, II. N NPN 2797 reiai? ?w 20s1 l
ZRA97018 August 29,1997                                                                                         I U.S. Nuclear Regulatory Commission Washington ~,.C. 20555 Attention:          Document Control Desk

==Subject:==
Zion Nuclear Power Station, Units 1 and 2 NUREG 0737, item II.D.1, Performance Testing of Relief and Safety Valves - Status of Open Issues NRC Docket Nos. 50-295 and 50-M

==References:==
: 1) Letter from C, Shiraki, NRC, to D. L. Farrar, Commonwealth Edison, dated October 6,1994, Safety Evaluation of Addtional Information Submitted in Response to Identifieu Deficienceis for Performance Testing of Safety and Relief Valves for Zion Nuclear Power Station, Units 1 and 2
: 2) Letter from J. H. Mueller, Comed, to U.S. Nuclear Regulatory Commission, dated September 13,1996, NUREG 0737, Item II.D.1 Performance Testing of Relief and Safety Valves
: 3) Letter from C. Shiraki, NRC, to I. Johnson, Commonwealth Edison dated October 4,1996, Performance Testing of Safety and Relief Valves - Zion Station, Units 1 and 2
: 4) Letter from T. W. Simpkin, Comed, to U.S. Nuclear Regulatory Commission, dated September 25,1995, NUREG 0737 Item II.D.1 Performance Testing of Relief and Safety Valves
: 5) Letter from C. Patel, NRC to II. Bliss, Commonwealth Edison, dated February 28,1989, NUREG-0737, Item II.D.1, Performance Testing          I of Relief and Safety Valves for Zion Station, Units 1 and 2 i  I This letter provides the current status of the outstanding NUREG 073'i, Item II.D.1 issues and provides new calculations associated with the Unit 2 Power Operated Relief Valves (PORVs).
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ZRA97018 Page 2 of 5 In a Safety Evaluation Re[$ ort (SER) dated October 6,1994, (Reference 1) the NRC Staffidentified three issues relative to the resolution of NUREG 0737. Item II.D.1, "Performano : Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valves." The issues are listed below with a discussion of the current status.
Issue (a) " Analyze the PORV discharge loads on the portion of piping upstream of the common header, and properly combine the PORV and PSV discharge loads with other loads including the SSE to verify the capability of the discharge piping to withstand the maximum expected loading."
Status       PORVs Based on plant design activities, new analyses associated with the PORVs have been completed. The items described in " Issue a" are addressed in Enclosure 1, " Pressurizer Power Operated Relief Valve Piping Hydrodynamic Loads" and Enclosure 2, " Piping Stress Analysis of Subsystem 2RC-02."
As a result of the enclosed calculations, field installation of modifications to Unit 2 pipe supports and reinforcement of main steel associated with one support have been completed to bring the Unit 2 PORV piping and support loads to within the allowable UFSAR design loads. Attachment A of this letter provides a sketch of the piping support changes. The enclosed calculations demonstrate the structural adequacy of the Unit 2 PORV discharge piping upstream of the common header by demonstrating that the UFSAR design limits are not exceeded. These calculations consider steam discharges at power, steam discharges followed by liquid discharges at power, and liquid discharges during low-temperature overpressure operation. The calculations combine the discharge loads with SSE loads.
Attachment N of the " load calculation" consists of microfiche of computer output files and is not included in Enclosure 1. The Tables and Figures contained in the body of the calculation and included attachments summarize the pertinent information from the output files. Comed considers the documentation provided in Enclosure I adequate to demonstrate acceptability of the methodology and the conclusions.
ki\liegroup\ misc-sub\porv\porvtran1. doc

ZRA97018 Page 3 of 5 Attachments M, N, and O of the " stress calculation" (Enclosure 2) are not included in Enclosure 2. These attachments consist of a large plot, diskettes of the computer model, and computer output, respectively. Again, Comed considers the documentation provided in Enclosure 2 adequate to demonstrate acceptability of the methodology and the conclusions.
Comed will perform a similar " stress calculation" for Unit 1. Note that the
                          " load calculation," Enclosure 1, applies to both Units. The Unit 1 " stress-    l calculation" and any required modifications will be completed prior to entering a Mode when the PORVs are required to be operable during the current Unit 1 outage. The Unit 1 " stress calculation" will be performed using the same methodologies and acceptance criteria as the Unit 2 calculation.
Pressurizer Safety Valves (PSVs)
Comed has evaluated the effects of the Unit 2 PORV piping support system modifications on previous PSV piping calculations to ensure continued operation is acceptable for the next operating cycle. These PSV piping calculations were previously reviewed and the results documented in an 1989 NRC Technical Evaluation Report (Reference 5). In that Technical Evaluation Report, the NRC did not indicate any disagreement with the overall calculation methodology, but did indicate that seismic loads should have been considered. In this evaluation seismic loads were combined with PSV discharge loads. The evaluation concluded that the Unit 2 PSV piping -
meets ASME Section III, Appendix F requirements and the piping system's functionality is not compromised. Furthermore, the overall conclusions of the previous calculations remain valid in that strain limits will be met and the integrity of the PSV piping will be maintained.                                  '
As described in Reference 2, Comed intends to perform plant modifications during the next refueling outages (ZlR16 and Z2R15) which will address PSV load combinations. In addition, the PSV modifications and associated analyses which will be completed in support of this modification will include PSV discharge loads in combination with SSE loads. In accordance with the NRC Staff request (Reference 3), Comed will provide a description of the PSV modifications for staff review and approval prior to installation.
Note that Reference 4 provided the basis for analyzing actuation of the PORVs and actuation of the PSVs separately.
kt\licgroup\rtisc-sub\porv\porvtrani. doc

ZRA97018 Page 4 of 5 issue (b) " Verify that the resulting maximum bending moments on the PSVs and PORVs are less than those demonstrated acceptable by testing."
Status      PORVs Comed's response (Reference 4) to the NRC SER issues (Reference 1) included information which indicated that the maximum bending moment applied to the PORVs would be less than the accepted EPRI test program values, and that the PORVs would be expected to operate satisfactorily. The enclosed " stress calculation" confirms that the bending moment applied to the Unit 2 PORVs remains less than the limits identified in the EPRI test       i program. The Unit I stress calculation will evaluate the bending moment applied to the Unit 1 PORVs against the limits identified in the EPRI test program.
PSVs The evaluation of the Unit 2 PORV piping support system modifications impact on the previous PSV calculations, described under " Issue (a),"
concluded that the resultant maximum bending moments on the Unit 2 PSVs are acceptable. The conclusion was based on the insignificant increase in the bending moments (< 0.02%) that were accepted in Section 4.3.1 of the Technical Evaluation Report (Reference 5). As a part of the PSV modifications committed to in Reference 2, the maximum applied bending moments will be determined and evaluated, i
Issue (c) " Assure that the maximum expected frictional and dynamic pressure losses for the inlet piping to the PSVs do not exceed that shown to be acceptable by testing."
Status      Comed's response (Referenw 4) to the NRC SER issues (Reference 1),
provided information which indicated that the maximum expected frictional and dynamic pressure losses for the inlet piping to the PSVs are less than the accepted EPRI test program results. The PSV modifications commited to in Reference 3 will be designed such that the expected inlet piping frictional    ,
and dynamic pressure losses for full flow steam conditions will result in      i stable performance of the PSVs.
k:\11cgroup\ misc-sub\porv\porvtran1. doc
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4 4
4 i ZRA97018
                      ' Page 5 of 5 Attachment B lists the commitments made by Comed in this submittal. - Please direct
                                                      ~
l                           any questions you may have concerning this submittal to this office.
i:                           Respectfully,
                      ":. H. Mueller L                           Site Vice President l                           Zion Station -
L Enclosures -
Attachments I
cc:                            . NRC Regional Administrator- RIII i
Mr. G. Grant - NRC RIII -
Zion Station Project Manager - NRR-
;                                                            Senior Resident inspector - Zion Station -
,                                                            Office of Nuclear Facility Safety - IDNS l=                                                           IDNS Resident Inspector Zion NLA
: j.                                                           Engineering Manager l                                                          - Master Files -
Reg. Assurance File DCD Licensing e
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