ML20202J329
| ML20202J329 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/17/1995 |
| From: | Wetterhahn M WINSTON & STRAWN |
| To: | Logan K NRC OFFICE OF INVESTIGATIONS (OI) |
| Shared Package | |
| ML20202J326 | List: |
| References | |
| FOIA-97-325 NUDOCS 9712110128 | |
| Download: ML20202J329 (205) | |
Text
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hir. Keith legan United States Nuclear Regulatory Commission Office of Investigations 475 Allendale Road King of Prussia, PA 19406 L
Dear hir. Logan:
In response to your request for specified documents n conjunction with your August 16, 1995 interview of Kenneth O'Gara, the following items are provided:
- 1. hiemora,.ium fwm 11. Herrick to F. Schnarr, "Non-conservatism in POPS Setpoint ATS Open item --Westinghouse NSAL PSE-93 204," hfEC-93 917. October 29, 1993
[ unsigned draft)
- 2. hiemorandum from II, Berrick to F. Schnarr, "Non-conservatism in P019 setpoint ATS open item -- Westinghouse NSAL PSE-93 204, h1EC-93 917.
December 30,1993
- 3. Letter from James C. Stone NRC Office of Nuclear Reactor Regulation to Steven E.
hiiltenberger, " Changes to Pressure-Temperature Limits in Technical Specifications. Salem Nuclear Generating Station, Unit 2," February 22,1994 [two copies, one with accompanying handwritten note]
- 4. Chronology of Licensing Involvement in Events Asseciated with POPS Non-conservatism Issue, April 1994
- 5. Discrepancy Evaluation Form DEF No. 94-0060, April 19,1994
- 6. Discrepancy Evaluation Form, DEF No. 94-0060 April 19,1994 [ copy with handwritten note]
9712110128 971200 PDR FOIA
,, , KEENAN97- 325 PDR f
hir. Keith Logan ^ August 17, 1995 Page 2
- 7. Nuclear Department incident Report Form, " POPS Setpoint Non-conservatism,"
(undated)
- 8. Nuclear Depanment Incident Report Fonn,
- POPS Setpoint Non-conservatism,"
(undated)
- 9. Nuclear Depanment incident Repon Form, " POPS Setpoint Non-conservatism,"
(undated)
- 10. Nuclear Depanment incident Repon Form, " POPS Setpoint Non-conservatism,
(undated)
- 11. Nuclear Department incident Repon Form, " POPS Setpoint Non-conservatism,"
(undated)
- 12. Nuclear Depanment incident Report Fonn,
- POPS Setpoint Non-conservatistr.," ,
(undated)
- 13. hiemorandum from 11. Herrick to J. Wiedemann, "Non-conservatism in POPS Setpoint Reopened as ATS Open item," h1EC-94-630, hiay 26,1994
- 14. hiemorandum from 11. Berrick to J. Wiedemann, "Non-conservatism in POPS
'Setpoint Reopened as ATS Open item," MEC 94-630, biay 26,1994 [ copy with margin notes]
- 15. Memorandum from K. hl. O'Gara to D. A. Smith, " Status of ATS Items While on Vacation," June 2,1994, with accompanying handwritten note
- 16. Memorandum from K. M. O'Gara to D. A. Smith, " Status of ATS Items While on
' Vacation," June 2,1994 [with margin notes]
- 17. Memorandum from D. A. Smith to J. Ranalli, "Non-conservatism in POPS Setpoint (DEF-94-0060) NLR 194400, September 28,1994
- 18. Nuclear Department' Incident Report Form, IR No. 94-419, November 17, 1994
- 19. Licensee Event Report No. 94-017-00, December 14' 1994, with accompanying letter ,
from J.- 11agan to U.S. Nuclear Regulatory Commission
. i hir, Keith legan August 17, 1995 Page 3
- 20. 10 C.F.R. 50.59 Safety Evaluation, December 21,1994
- 21. 10 C.F.R. 50.59 Safety Evaluation, February 6,1995
- 22. hiemorandum from K. O'Gara to D. Smith, " Evaluation of Reporting Requirements and Root Cause Assessment Associated with POPS Issues (IR 95-343)," (draft)
April 6,1995
- 23. hiemorandum from K. O'Gara to D. Smith, " Evaluation of Reponing Requirements and Root Cause Assessment Associated with POPS Issues (IR 95 343)," (draft) hiay 16,1995
- 24. Licensee Event Report No. 94-017 01 (draft), August _,1995, with cover telecopy page from W. IIall to K. O'Gara
- 25. Licensee Event Report No. 94-017-01 (draft) August , 1995
- 26. Letter from J. llagan to U.S. Nuclear Regulatory Commission, (draft) "AShiE Code Case N-514, Salem Generating Station Unit Nos. I and 2. NLR N94193 (draft)
(undated) (copy with handwritten note]
- 27. Letter from J. IIagan to U.S. Nuclear Regulatory Commission, (draft) " ash 1E Code Case N 514, Salem Generating Statica Unit Nos. I and 2. NLR N94193 (draft)
(undated) [ copy with margin notes)
- 28. hiemorandum from D. A. Smith to T. K. Ross, Use of Pressurizer PORVS To hiitigate inadvertent Si at Power Transient (draft) NLR-194553 (undated)
- 29. Telecopy page from ht. Pastva to D. Stri. .. '.pril'5,1995, with attached Nuclear 3
Department Incident Report form 95 343, April 4,1995, and accompanying handwritten note
- 30. Telecopy dated October 4,1994 with accompanying " Attachment to Problem Report No. 940927126," (undated)
- 31. LER Response input (draft) December 6,1994
- 32. LER Response input (draft), undated y
4 hir Keith logan AuEust 17, 1995 Page 4
- 33. LER Respons- Input (draft), undated
- 34. LER Response input (draft), undated
- 35. Four Week Look Ahead, Kenneth M. O'Gara, January 17,1994
- 36. Four Week Look Ahead, Kenneth hl. O'Gara, January 31,1994 l 37. Four Week Look Ahead, Kenneth bl. O'Gara, January 31,1994 (with margin note l " update"]*
- 38. Four Week Look Ahead, Kenneth hl. O'Gara, February 7,1994
- 39. Four Week Look Ahead, Kenneth ht O'Gara, February 14,1994
- 40. Four Week Look Ahead, Kenneth hl. O'Gara. February 21,1994
- 41. Four Week look Ahead, Kenneth hl O'Gara, February 28,1994
- 42. Four Week Look Ahead, Kenneth hi O'Gara, N1 arch 7,1994
- 43. Four Week Look Ahead, Kenneth hl. O'Gara, hlarch 14,1994
- 44. Four Week Look Ahead, Kenneth hl. O'Gara, March 21,1994
- 45. Four Week Look Ahead, Kenneth h1. O'Gara. h1 arch 28,1994
- 46. Four Week Look Ahead, Kenneth hi. O'Gara, April 11,1994
- 47. Four Week Look Ahead, Kenneth hl. O'Gara, April 18,1994
- 48. Four Week Look Ahead, Kenneth hl. O'Gara, hiay 2,1994 49 Four Week Look Ahead, Kenneth M. O'Gara, May 9,1994
- 50. Four Week Look Ahead , Kenneth M. O'Gara, May 16,1994 Referenced by handwritten date.
Mr. Kehh Logan August 17, 1995 Page 5
- 51. Four Week Look Ahead, Kenneth M. O'Gara, May 31,1994
- 52. Four Week Look Ahead, Kenneth M. O'Gara, June 27,1994
- 53. Four Week Look Ahead, Kenneth M. O'Gara, July 5,1994
- 54. Four Week Look Ahead, Kenneth M. O'Gara, July 11,1994
- 55. Four Week Look Ahead , Kenneth M. O'Gara, July 18,1994
- 56. Four Week Look Ahead, Kenneth M. O'Gara, July 25,1994
- 57. Four Week Look Ahead, Kenneth M. O'Gara August 1,1994
- 58. Four Week Look Ahead, Kenneth M. O'Gara, August 8,1994
- 59. Four Week Look Ahead, Kenneth M. O'Gara, August 15,1994*
- 60. Four Week Look Ahead, Kenneth M. O'Gara, August 29,1994
- 61. Four Week Look Ahead, Kenneth M. O'Gara, September 6,1994
- 62. Four Week Look Ahead , Kenneth M. O'Gara, September 26,1994
- 63. Four Week Look Ahead, Kenneth M. O'Gara, October 3,1994
- 64. Four Week Look Ahead, Kenneth M. O'Gara, October 17,1994
- 65. Four Week Loo.. . icad , Kenneth M. O'Gara, October 24,1994
- 66. Four Week Look Ahead, Kenneth M. O'Gara, October 31,1994
- 67. Four Week Look Ahead, Kenneth M. O'Gara, November 14,1994
- 68. Four Week Look Ahead. Kenneth M. O'Gara, November 28,1994
'T Referenced by handwritten date.
\
c !
* , i 1
hir. Keith Logan August 17, 1995 Page 6
- 69. Four Week Look Ahead, Kenneth hl O'Gara, December 12,1994.
- 70. Four Week Look Ahead , Kenneth ht. O'Gara, December 19,1994
- 71. Four Week Look Ahead, Kenneth bl. O'Gara, January 23,1995
- 72. Four Week Look Ahead, Kenneth bl. O'Gara, January 30,1995
- 73. Four Week look Ahead, K(.tneth hl. O'Gara, h1 arch 6,1995
- 74. - Four Week Look Ahead, Kenneth hl. O'Gara, hiarch 13,1995 The Final Collegial Self-Assessment Report you had requested will be sent to you under separate cover, as we are evaluating whether it is a document that would merit confidential treatment under 10 C.F.R. 2.790.
If you require further infonnation in conjunction with these documents, or if I can be of further assistance, please let me know, Sincerely, i , f h i/L j f(;(T(l l:a!'a .i . '? hlark J. \ etterhahn - Counsel for Public Service Electric & Gas Company hijW/ayc Enclosures l
- . O PSIKi Pubhc Service Electne and Gas Company P O Box 2M Hancocks Brege. New Jersey 08038 Nuclw Deputment MEC-93-917 TO
- F. Schnarr Reliability & Assessment Grcup FROM: Howard Berrick Salem Mechanical Engineering Supervisor
SUBJECT:
NONCONSERVATION IN POPS SETPOINT ATS OPEN ITEM - WESTINGHOUSE NUCLEAR SAFETY ADVISORY LETTER PSE-93-204 (NSAL-93-005B) DATE: December 30, 1993
Background
Westinghouse NSAL-93-005B transmitted via PSE-93-204 identified a potential issue regarding a nonconservatisim in the POPS setpoint develcpment. The pressure difference from the wide range pressure transmitter to the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature limits are defined) was not considered in Westinghouse an& lysis. This pressure difference effectively results in the pressure in the-reactor vessel midplane being greater than that seen by the wide range pressure transmitters used to actuate the PORVs, potentially resulting in violation of the Tech. Spec heatup and ecoldown pressure / temperature limit curves. The Salem POPS analysis (SGS/M-DM-042 and 062) used methodology provided by Westinghouse in their report " Pressure Mitigating ' Systems Transient Analysis" (July 1977). The methodology in this report did not consider the pressure difference of concern and thereforte the subject NSAL applies to Salem 1 & 2. Discussion The Tech. Spec. heatup and cooldown P/T limit curves (attached) are determined in accordance witn the requirements of Appendix G of 10CFR50 and ensure reactor vessel integrity. The pressuriter overpressure protection system (POPS) protects the RCS from exceeding the Tech. Spec. P/T limit curves by opening the PORVs during cold overpressure transients (RCS below 312*F). IF W n vt.r Itt"i. . A ._.. .
' F. Schnarr 2 12/30/33 i i The POP $ uses the two wide-range RCS pressure sensors PT403 and PT405 to actuate the r0RVs. These sensors sense hot leg pritssure. The pressure at the vessel midplane will be higher than the pressure at the hot leg due to the dynamic and static pstessure difference between the locations. The dynamic pressure difference depends on the number of reactor coolant pumps (RCPs) in operation at the time. Westinghouse did not consider the delta-P associated with the difference in location of the wide-range transmitter relative to the vessel midplane leading to a concern that the POPS setpoint may be nonconservative. The issue of POPS setpoint for actuating the PORVs must be shown to provide adequate protection, with the additional delta-P incorporated in the setpoint analysis. The Salem POPS analysis calculated the maximum pressure attained during a cold overpressure transient to be 446 psig with the PORV set at 375 psig. Therefore, it must be shown that 446 psig plus the delta-P of concern does not exceed the Tech. Spec. P/T limits, in order for the POPS PORV setpoint to be adequate. The Tech. Spec. P/T limit curves define the allowable temperature and pressure combinations for heatup rates up to 60' F/hr and far cooldown rates ranging from O'F/hr to 100*F/hr. For the POPS analysis a composite curve made up of the heatup curve and the 20*F/hr cooldewn curve is used. The use of cooldown rate of 20*F/hr is 3ustified because at the low temperatures when POPS is armed, higher cooldown rates are not achievable. A review of the Tech. Spec. P/T limit curves shows that the 20'F/hr cooldown curve is more limiting at low temperatures, on both Units. The pressure that must not be exceeded is 450 psig on Unit 1 and 475 psig on Unit 2. Additional margin in the Tech Spec. curves can be gained for the POPS application by taking credit for ASME Code Case N514. This code case states "LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of Section XI, Article G-2215". (LTOP - Low By Tempressure Overpressure Protection is the same as POPS). taking credit for this Code case, the allowable pressure can be increased by 10%. In this case the lowest pressure that must not be exceeded is 495 psig on Unit 1 and 522.5 psig on Unit 2. b '
r-F. Schnarr 3 12/30/93 4 Evaluation Table 1 summarizes the r.sults of the evaluation. Comparing the POPS analysis maximum pressure of 446 psig to the Tech. Spec. P/T li. Tits shows that the margin available to accommodate the delta-P is 4 psig on Unit 1 and 29 psi on Uni'; 2. An additional 45 psi (Unit 1) and 47.5 (Unit 2) can be gained by taking credit for the code case. Westinghouse indicated that based on generic analyses the delta-P is 74 psi with four RCPs operating. It is clear that Unit 1 does not currently have the margin to accommodate the expected delta-P with four pumps operating. To quantify the Salem specific delta-P and assess benefit of fewer operating pumps, Westinghouse was requested to calculate the delta-P for one, two and four RCPs operating. The results of the calculation provided delta-P values of 31 psi, 39 psi and 73 psi for one, two and four RCTs respectively (PSE-93-707). Westinghouse assumed the transmitters are zerced out to the RHR suction line at 92.4 ft. The transmitters are zereed to the hot-leg (97 ft). To correct for this difference 2 psi was added to the Westinghouse results. The maximum pressure including the above delta-P values is presented in Table 1. Table 1 shows that Unit 1 Tech. Spec. minimum of 450 psig is exceeded by the two RCPs and four RCP cases. Taking credit for the Code case, the Unit 1 Te:h. Spec. minimum is 495 psig which can be met by the two RCPs operating case. The Tech. Spec. pressure limit increases with increasing temperature and exceeds the four RCPs maximum pressure of 517 psig at 200* F. Therefore, by restricting the number of operating RCPs to two RCPs below 200' F, the POPS PORV setpoint will provide adequate protection. The temperature of 200* F was selected to coincide with cold shutdown Tech. Spec. operational mode (Mode 5) . Table 1 shows that Unit 2 Tech. Spec. minimum press'1re of 522 psig (taking credit for the Code case) can be met with four RCPs operating. However, to maintain similarity in the operation of the units, and to provide margin for future evaluations of~the Tech. Spec P/T curves, the same restriction on RCP operation is recommended on Unit 2. It should be noted that 1) restricting the number of RCPs is one of the recommendations in the subject NSAL and 2) taking credit for the 10% margin in the limits as afforded by the ASME Code case was discussed with Westinghouse and this margin has been credited by other utilities (eg FP&L), to address the subject issue.
%e
s -
-l F. Schnorr 4 13/30/93 l
(. Recommendation i In, summary to address the POPS setpoint nonconservatisms-
. identified in Westinghouse Nuclear Safety Advisory letter PSE -
204, we recommend restricting the nunter of RCPs in operation while-in-mode 5 to no more than two RCPs. Procedure change - request is being issued to incorporate this change into IOP - (Cold Shutdown to Hot Standby) and IOP-6 (Hot Standby.to Cold Shutdown). _The ATS open item NSAL-PSE-93-204 is considered closed by this letter.. GN: Attachment-
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C H. Dansk ' (/ K. Pike-
~J..Ranalli J.'Serwan- l J. Wiedemann ATS File MEC File Standards Records'Ccorindator 9
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a* . TABLE 1 POPS. jMaximum Delta P Maximum Tech Specs Tech Spec. PORV Pressure (Vessel Pressure + P/T limits Setpt P/T limits Calculated midplane to Delta-P minimum minimum Psig. In POPS transmitter pressure pressure +10% Analysis psig Psig 4 RCPs 2 RCPs 4 2 (psi) (pel) RCPs RCPs Psig psig Unit 1 375 446 73 39 519 485 450 495 Unit 2 375 446 73 39 - 519 485 475 522.5 x
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g e - l' . g W, l Chronoloav of Licensina Involvement in Ev nts_-
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Associated with POPS NoncgDservatista Issu M], 1994 Licensing and Regulation reviewed the results of Mechanical Engineering's (ME) evaluation of the subject de iciency documented 41nimemo MEC-93-917 dated December.30, 1993. This memo-was< brought to the attention of L&R by the Salem Tech. System Engineer, this memo evaluated:a potential non conservatism in-L the development;of the Precsurizer Overpressure Protection: System 4 L (POPS)-Setpoint that was identified:to PSE&G by thi NSSS_ vendor i .(NSAL-93-005B). Specifically, that the differential _ pressure ! between the midplanesof the reactor vessel and:the location et the pressure transmitters which signal the POPS.valver to open was not-considered in'the original analyses. The ME evaluation concluded that-with 2 RCPs operating the Pressure / Temperature (P/T). limits; contained in-the Technical Specifications would be-exceeded. However, ME credited ASME Code Case N-514 which' allows the P/T limits to be exceeded by 10% such that sufficient margin would exist to address.the differential pressure due to the two
. RCPs-operating. The specific details of this; evaluation are documented in Attachment 1.
After L&R-review, conversations with ME, the system Engineer, and Principal Engineer .L&R, L&R requested that ME issue _a
- Discrepancy Evaluation Form (DEF)-and evaluate--the-issue ~to
-determine if any cperability concerns existed. The DEF was written e. April 19.1994 -(Attachment 2)
L&R reviewed the contents of the DEF, and based on.the information that was-known at - that' time, believed that a potentially reportable condition did exist for:the follo ng reasons: 6 Credit for ASME Code Case.N-514 allowed 10% margin,.but the use lof this code case'had not been approved.for Salem.; -Without the cCode' Case Margin, the P/T. limits.would be exceeded.
- RH3;would also be available to provide overpressure protection however, plant specific analysis had not been completed.
Based on'the DEF conclusions and the information above, L&R drafted an Incident Report (IR) Form on April 21, 1994
-(Attachment 3)-indicating that although it did not appear to be
-an immediate. operability or safety concern, the issue should be considered reportable .iri accordance with 10CFR50.72 since the
-POPS.was outside1the design basis.
>A meeting was held on this day between ME, Engineering Sciences &
L&R to_. discuss 1the contents of the draft IR and whether a 1(
4 9 Chronoloav of Licensinc Involvement in Events Associated with POPS Nonconservatism Issue (Cont'd) reportable deficiency existed. It was determined that sufficient information was not available at this time, and additional Engineering evaluation was warranted based on the following the use of valve RH3 and if the valve could be considered part of the overpressure protection design basis. Ortrating precedure controls that prohibit the operation of the RCPs waen the pressurizer is water solid (i.e. , procedures required that a bubble be established that would mir.imize overpressure tran_ients). If this was procedurally
- controlled, it may be valic not to consider the differential
! pressure associated with RCPs running and the P/T limits ! would not be exceeded Engineering agreed to further evaluate the issue considering current operating procedures, and the draft IR was put on hold pending completion of this evaluation. It is noted that the draft IR was revised after.this meeting to add additional infc'"ation regarding RH3 and removal of the ACI. Again, this was craft and never issued. May 26, 1994 Engineering completed the reevaluation of the NSAL as documented > in Memo MEC-94-630 dated May 26, 1994 (Attachment 4). Engineering concluded that by restricting number of RCPs to one, _ and assuming no RCPs operating.When the pressurizer is water solid, the current P/T limits would be met with a minor exceedance (0.7 psi) assuming only 1 POPS valve is available. Engineering dispositioned this as currently within the design basis for POPS based on recent informal _ calculations using the GOTM7C Code. Therefore, an IR was not warranted and the DEF was considered closed. The following two actions were identified in the Engineering Response to provide additional margin on the P/T curves:
- 1) Submit _ Request to NRC for approval of ASME Code case N-514,
- 2) Prepare License Change Request (LCR) to add RH3 to uhe POPS Technical Specifications for both Units.
i (
6 6 Chronoloov of Licensina Involvement in Events Associated with POPS Nenconservatism Issue (Cont'd) The margin on the 9/T curves could be further reduced in the future as the vessel becomes further irradiated. Therefore, L&R ' was requested to pursue both of these actions in the near term. Following the issuance of the May 26 memo., a meeting was held between L&R, ME, Engrg Sciences and Salem Tech. to discuss the results of the evaluation. During this meeting, ME again clarified that the 0.7 psi exceedance is accepteble based on the way the actual pressure limits were interpreted right from the curves contained in the plant Tech. Specs. The curves make it extremely difficult to accurately determine the pressure limit ' from, even to an accuracy of within 2-3 psi. Therefore, it was agreed that the .7 psi exceedance was not significant. June 13, 1994 Engineering completed calculation S-C-RC-MDC-1358 to support the addition of valve RH3 to the POPS Technical Specification. This calculation (Attached to Letter NLR-N94193 dated 12/22/94) was provided to L&R to use in the LCR development recommended by ME in the May 26 memo. Key assumptions of this calculation was that either 2 POPS or 1 POPS valve and RH3 would provide adequate relieving capability in the event of an overpressure transient. This calculation was performed using the GOTHIC computer code. Aucust, 1994
-L&R began preparation of a letter to the NRC requesting approval of Code Case for Salem, and also began preparation of a LCR related to the P/T limit curves to include this information in a document outside of the Tech. Specs., This included adding RH3 to the POPS Tech. Spece..
During the preparation of these letters to the NRC, L&R raised several questions related to the calculation S-C-RC-MDC-1358. These issues were informally discussed on a few occasions with ME and Engineering Sciences. This eventually resulted in a meeting between the System Engineer, L&R, ME-and Engineering Sciences on September 16, 1994. Concerns raised during this meeting by L&R are documented in memo NLR-194400 to ME and Engineering Sciences on 09/28/94 (Attachment 5) In summary, the following L&R concerns were discussed:
- 1) The calculation conservatively assumed that the pressurizer would become water solid following the inadvertent start of a Intermediate Head SI pump, and added the pressure differentici associated with one or more RCPs operating.
However, the analyses to address the DEF (Memo MEC-94-630 dated May 26, 1994) did not assume any RCPs operating. The potential existed that the pressurizer would eventually become water solid in a period of time following start of an 3
s Chronoloov of Licensina Involvement in Events Associated with POPS Nonconservatism Issue (Cont'd) Intermediate Head SI pump when the POPS lif t and the bubble is relieved, with a RCP(s) operating. Therefore, the differential pressure should be included in analyses documented in MEC-94-630. If this was the case, then the P/T limits would be exceeded.
- 2) Assess the ability to assume the High Head (centrifugal Charging) SI Pump flow rate based on administrative controls to remove power from the Intermediate Head SI pump. ;
- 3) Issue a Problem Report documenting the results of the additional analyses based on the assumptions discussed in item 1 and 2 above.
The problem report was ijsued on 9/27/94 documenting the results of the reanalysis (Attachment 6). The reanalysis concluded that the current Tech. Spec. Bases for POPS (1 POPS provides sufficient relief eapacity to ensure the P/T limits are not exceeded) is met for the start of a High Head SI Fump with 1 RCP operating. This is consistent with current operating procedures that limit the availability of the Intermediate Head SI Pump during Mode 5. At this time this issue was considered closed, and L&R again began preparation of a Letter to the NRC requesting approval of the use of ASME Code Case N-514. November, 1991 ASME Code Case Letter was in signout. The Manager - Engineering Sciences raised a question with regard to operation of the PDP in Mode 5. Should an inadvertent SI signal be generated, the PDP would continue to operate if offsite power is available. 'This would result in an overpressure transient where the combined flow of a High Head SI Pump and the PDP would now be considered the worst case mass addition transient. Engineering Sciences reevaluated the POPS analysis and concluded that the Unit 1 F/T limits would be exceeded. Based on the above, L&R generated an IR for POPS being outside the design bases. This was reported to the NRC as LER 27.'/94-017 dated November 17, 1994 (Attachment 6). Unit 2 was still within the deoign bases. December, 1! '4 LER 272/94-017 dated December 14, 1994 was submitted to the NRC summarizing the event. On December 16, 1994, a Letter was issued to the NRC outlinirg the administrative controls to be used to ensure the P/T limits for both Units would be maintained during Mode 5 operation (Attachment 7). On December 22, 1994,- PSE&G submitted to the NRC a Request for ASME Code Case N-514 (Attachment 7).
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Page 1 of _ WUCLEAR COMMON Page 1 of 4 Revision 4
0 e CONBENTS ATTACITFD TO RCS DEF:
-[1]AS IDENTIFIED IN THE MEC-93 917, THE CALCULATED MAXIMUM PRESSURES AGAINST THE APPENDIX G ALLOWABLE FRESSURE ARE AS FOLLOWS FOR SALEM 1&2.
UNIT RCP IN CALCULATED MAX. TECH. SPEC P/T SERVICE PSEG PSEG 1 2 485 450 1 1 477 450 2 2 485 475 2 1 477 475 (2) AS IDENTIFIED IN THE MEC-93-917, ADDITIONAL MARGIN IN THE TECH SPEC CURVES CAN BE GAINED FCR THE LTOP APPLICATION BY TAKING CREDIT FOR ASME CODE CASE N514. THIS CODE CASE STATES "LTOP SYSTEMS SHALL LIMIT THE MAXIMUM PRESSURE IN THE VESSEL TO 110% OF THE PRESSURE DETERMINED TO SATISFY APPENDIX G OF SECTION XI, ARTICLE G 2215". BY TAKING CREDIT FOR THIS CODE CASE, THE ALLOWABLE PRESSURE CAN BE INOREASED BY'10 %. IN THIS CASE THE LOWEST PRESSURE THAT MUST NOT EXCEED IS 495 PSIG FOR SALEM 1 AND 522.5 PSIG ON UNIT 2. THIS WILL ELIMINATE THIS DEF DISCREPANCY, IF APPROVED..USE OF THE CODE CASE HAS BEEN RELIED UPON BY OTHER UTILITY [FP
&L) . HOWEVER, USE OF CODE CASE WILL REQUIRE NRC PERMISSION AND POSSIBLE REVISION OF TECH SPEC CURVES TO ADDRESS 10 %
INCREASE. [3] LIMITING THE RCP OPERATION TO NO MORE THAN 2 RCPS IN ,- MODE 5 IS ALREADY INCLUDED IN THE CURRENT PLANT PROCEDURES. ,, g ' (THROUGH IMPLEMENTATION OF PROCEDURE REVISION REQUEST: , R07326). PROCEDURE CHANOE TO LIMIT RCP OPERATION TO ONE PUMP ' f IN OPERATION IS BEING PURSUED BY THE SYSTEM ENGINEER. (4) THE CURRENT PLANT DESIGN RELES ON ONE PORV SET AT 375 PSIG. [5] ORIGINAL RHR DESIGN AT SALEM INCLUDED SUTO CLOSURE (i.CI) INTERLOCK OF RH 1 & 2 VALVES TIED TO PT403 AND PT405 PRESSURE > 375 PSIG. SALEM UNITS REMOVED THESE ACI THROUGH DCPS. BASED ON WESTINGHOUSE WCAP 11640, IF THE ACI IS REMOVED, THEN THE INADVERTENT ISOLATION OF THE RHR RELIEF VALVE IS CONSIDERED TO BE HIGHLY UNLIKELY. RER RELIEF VALVES ARE THEN AVAILABLE TO MITIGATE POTENTIAL LOW TEMPERATURE OVER PRESSURE TRANSIENTS. THE RELIEF VALVE SET POINT OF SALEM UNITS ALONG WITH THE RH3 VALVE CAPACITY WERE
-GENERICALLY EVALUATED BY WOG TO BE ADEQUATE TO PROVIDE APPENDIX G PROTECTION WITHOUT RELYING ON THE USE OF PORVS FOR LTOP. ALTHOUGH THE PLANT SPECIFIC ANALYSIS FOR LTOP SYSTEM USING RH3 RELIEF VALVE IS NOT COMPLETED YET, THE RESULTS ARE EXPECTED TO PRODUCE PEAK PRESSURE WITHIN THE 10
- 4. e
U
% ACCUMULATION OF THE SET PRESSURE. ADDITION OF PRESSURE DIFFERENCE BASED ON ONE RCP WILL PRODUCE ACCEPTABLE RESULTS. (INITIAL ANALYSIS COMPLETED TAKING CREDIT FOR RH3 PROVIDES PEAK PRESSURE FOR MASS INPUT CASE SLCH THAT THE CURRENT TECH SPECS CAN BE MET WITHOUT THE USE OF CODE CASE.
THE HEAT INPUT CASE HAS NOT BEEN COMPLETED, BUT IS LIKELY TO BE NON LIMITING CONSIDERING THE CURRENT DESIGN FOR WHICH THEl HEAT INPUT CASE IS NOT LIMITING.) . M) UPON COMPLETION OF THE RH3 LTOP CALCULATION, THE SALEM FSAR WILL HAVE TO BE REVISED THROUGH A 10CFR 50.59 EVALUATION. TECH SPEC REVISION IS NOT EXPECTED, BUT IF MANDATED, LCR WILL HAVE TO BE GENERATED. [7] THE DISCREPANCY IDENTIFIED IN THE DEF CAN BE RESOLVED THROUGH ONE OF THE FOLLOWING APPROACHES: (A) GETTING ASME CODE CASE N514 APPROVED BY NRC FOR SALEM. INCIDENTALLY, THE CODE CASE HAS NOW BEEN INCLUDED IN THE ASME XI THROUGH 1993~ ADDENDUM. OR (B) COMPLETING THE CALCULATION FOR LTOP3.'UPI REVISING PLANT DESIGN BASIS. r [8] THIS DEF IS NOT CONSIDERED AN OPPRABILITY CONCERN TAKING y$ THE CREDIT FOR CODE CASE 514 OR THE 1993 ADDENDUM OF ASME/ ./ - XI.
m FORM NC.MR-AF.55-0006-1 I e NUCLEAR CEPARTMENT INCIDENT RSPORT FORM i l COMMITMENT NUMBER INCIDENT REPORT No. USE CONTINUATION SHIETS IF NECESSARY SECTION I REPORT
SUBJECT:
(Initiator)PnRs K A s4 Alom /b u .;m /:< m UNIT S3,HC): DATE OF INCIDENT: 4 / 20 /,ff, TIME: .,
SUMMARY
OF EVENT (IF ESF ACTUATION, INCLUDE SOE PRINTOUT): Pleau. se, R6vLA f . REPORTED BY K /)h DEPT: M.A PHONE EXT L' 70 SECTION II (SNSS/ OPS MG) RX PWR AT TIME OF EVENT: 4 UNIT LOAD: MWe Op Con / Mode REPORT MADE PER ECG7 (Y/N): (I, YES, ATTACH ECG COPY) LCO #: A/S #: DATE IN: TIME IN: W.R.#: INITIAL CAUSE DETERMINATION: EQUIP DESIGN PERSONNEL PROCEDURAL OTMER: REPORTABLE: YES/NO, REASON SNSS/NSS SIGNATURE: DATE: / / COMMENTS: OPERATIONS MANAGER REVIEW : _ DATE: / / _ _ - N=I..r Co-o. .. . I or 4 R.v. s
t 2 The supplemental response to GL 92-( for Hope Creek on Fluence Levels is with you for your signature. This needs to go to SLB
- and should be given some priority.
I have ctill been tr:'ing to meet with Pete Ott to discuss the Charging Pump LCR. He cancelled the last meeting we had arranged and was working the back chift last week. This week he's been in Pittsburgh. The folder is on my desk. Maybe h. Villar could
.teet with him to--finalize the change. If not,_I'll resolve it when I return.
. ' 11 also take care of the POPS setpoint issue when I return.
However, I would still like to discuss the open issues with H. Berrick before I leave (i.e., 450 vs. 450 7 max. pressure l following an inadvertent SI, and the need to perform an analysis of the Inadvertent SI, with a bubble in the pressurizer and 1 RCP in service). If you need to reach me, I'll be unavailable. Only Kidding. See you in a coup'e of weeks. i
/ 4.[ ~. .
/C 8
'(
Pu%c Service Electnc and Gas Company P O. Box 236 Vancocks Bridge New Jersey 08038 Nuclear Department MEC-94-630 TO: J. Wiedemann Technical Engineer e i FROM: H. Berrick Salem Mechanical En inee f Sup rvisor
SUBJECT:
NONCONSERVATISM IN POPS SETPOINT 4 REOPENED ATS OPEN ITEM - WESTINGHOUSE NUCLEAR SAFETY { ADVIGORY LETTER PSE-93-204 (NEAL-93-005B)
REFERENCE:
MEC-93-917 Dated 12/30/93 from Berrick to Schnarr DATE: May 26, 1994
Background
Westinghouse NSAL-93-005B transmitted via PSE-93-204 identified a potential issue regarding a nonconservatism in the POPS setpoint development. MLC-93-917 referenced above provided the resolution. This memorandum (MEC-94-630) supersedes MEC-93-917 The pressure difference from the wide range pressure transmitter to the reactor vessel mid plane (where the Tech. Spec heatup and coolduwn pressure / temperature limits are defined) was not considered in West.inghouse analysis. This pressure difference effectively results in the pressure in the reactor vessel mid plane being greater than that seen by the wide range pressure transmitters used to actuate the PORVs, with a potential to result in violation of the Tech. Spec. heatup and cooldown pressure / temperature limit curves. The Salem POPS analysis (SGS/M-DM-042 and 062) used methodology provided by Westinghouse in their report " Pressure Mitigating Systems Transient Analysis" (July 197'l). The methodology in this report did not consider the pressure difference of concern and therefore the subject NSAL applies to Salem 1 & 2. p\4
- n. -a.s '
H J'H ag,
,- . J. Micd:monn 2 5/26/94 l
l Discussign The Tech. Spec. hestup and cooldown P/T 11mit curves (attached) are determined in accordance with the requirements of Appendix G of 10CFR50 and _ ensure reactor vessel integrity. The pressurizer overpressure protection system (POPS) protects the RCS from exceeding the fech.-Spec. P/T-limit curves by opening the PORVs during cold overpressure transients (RCS cold leg temperature below 312
- F) .
The POPS uses the two wide-range RCS pressure sensors PT 403 and PT 405 to actuate the PORVs. These sensors sense hot leg pressure. The pressure at the vessel mid plane will be higher than the pressure at the hot leg due to the dynamic and static pressure difference between the locations. The dynamic pressure cifference depends on the number of reactor coolant pumps (RCPs) in operation at the time. Westinghouse did not consider the delta-P associated with the difference in location of the wide-range transmitter _ relative to the vessel mid plane leading to a concern rhat the POPS setpoint may be non conservative. The Salem POPS analysis (SGS/M-DM-042 and 062) ca: culated the maximum pressure attained during a cold overpressure transient to be 446 psig for the mass input case and 418 psig for the-heat input case with the PORV set at 375 psig. Therefore, it must be shown that the above peak pressures, plus the additive pressure i ' based on the Westinghouse notification'does not' exceed the Tech.. Spec. P/T limits, in order to comply with the Appendix G requirement. The Tech Spec. P/T limit curves define the allowable temperature a and pressure combinations for heatup rates up to 60'F/hr and for cooldown rates ranging trem O'F/hr to 100*F/hr. For the POPS analysis a composite curve made up of the heatup curve and the 20*F/hr cooldown curve is.used. The use of cooldown rate of 20*F/hr is justified because'et the low temperatures when POPS is armed, higher cooldown rates are not achievable. A review of the Tech. Spec. P/T limit curves shows that the 20*F/hr cooldown curve is more limiting than heatup curve at low temperatures, on both Units. The pressure _that must not be exceeded is 450 psig on Unit 1 and 475 psig on Unit 2. k
J. Wicdtmann 3 5/26/94 Additional margin in the Tech Spec. curves can be gained for the POPS application by taking credit for ASME Code Case-N514. This code case states "LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of Saction XI, Article-G-2215". (LTOP - Low Temperature-Overpressure Protection is the same as POPS). This Code Case has been now incorporated in the 93 Winter Addenda of ASME Section XI. By taking credit for this Code case or the 1993 Code,-the allowable pressure can be. increased-by 10%. In this case.the lowest pressure that must not be exceeded is 495 (vs 450] psig on Unit 1 and 522.5 (vs 475) psig on Unit 2. However, neither the Code Case or the updated version of the Code can be applied at this. time pending NRC approval of them for Salem. Evaluation Tables 1 & 2 summarize the results of the evaluation. Comparing the POPS analysis maximum pressure of 446 psig to the Tech. Spec. P/T limits shows that the margin available to accommodate the delta-P ic 4 psid on Unit 1 and 23 psid on Unit 2. l To quantify the Salem specific delta-P and assess the cenefit of
- fewer operating pumps, Westinghouse was requested to calculate the delta-P for one, two and four RCPs operating. The results of the calculation provided delta-P values of 29 psi, 37 psi and 71 psi for one, two and four RCPs respectively (PSE-93-707).
Westinghouse assumed the transmittars are zeroud out to the RHR suction line at 92.4 ft. The transiitters are zerced to the hot-leg (97 ft). To correct for this. difference 2 psi was added to the Westinghouse results. The maximum pressure including the above delta-P values or just the static pressure difference between the transmitter and the cora mid plane, as applicable, is presented in Tables 1 & 2. Table 1 shows that Unit 1 Tech. Spec, minimum of 450 psig is exceeded by less than 1 psi for the inadvertent start of one SI pump for the mass input case. This minor exceedance of the heat up and cooldown curve is not considered an infringement of the Appendix G concern for the following reasons. [a] Recent informal calculation using GOTHIC has reestimated peak pressure for mass input case to be 438 psig, and (b) Recent calculation usino RH 3 valve to provide LTOP mitigation has calculated peak pressure for mass input case as 420 psig (Salem has removed Autoclose Interlock).
, J. Wicdemenn 4 5/26/94 j-Recommendation In summary to address the POPS setpoint non conservatisms identified.in Westinghouse Nuclear Safety. Advisory letter PSE 204, we recommend restricting the number of RCPs in operation while in Mode 5.to no more than'one RCP. This has been already incorporated into station procedure IOP-2 (Cold-Shutdown to Hot Standby) and IOP-6-(Hot Standby to Cold Shutdown). Future changes to the. Station procedures should not rescind the restriction of Mode 5 operation with no more than one RCP. The next capsule-on Salem Unit 1-is scheduled for removal during the Spring 1995 refueling outage. To address any nonconservative shift of the Appendix G curves at that time, (1) Licensing should pursue approval of Code Case 514 for Salem ano (2) Initiate a License Change Request to take credit for RH3 safety relief valve for LTOP. This ATS open item NSAL-PSE-93-204 is considered closed by this letter.
'(A4RD:
Attachments y c-C M. Danak V. Chandra C. Lashkari K. O'Gara K. Pike J. Ranalli J. Serwin F. Schnarr D. Smith ATS File MEC File Standards Records Coordinator l __.J
TABLE.1-I-
SALEM ESTIMATED INITIATING' EVENT DELTA
lteet ' . e.e Gospettes - ie / : 3 - e ' y 784 9 M0 d ' M L laserets 1 Acessantle ugeessetta , , < ' 8eerettee fast fase. 284 (33F9) = ' me Servie 8erees W - 18 597 0 ^ O 98 11 4 ' ' 11 4' agi W an' Jag 400 ele too 18G444f88 FEustaafvet (OSS.F) t CINTA!E W tt4 MIN FM MES!ILE 116thself costias rigene 3.4-3 . Sales Unit 1 Neester Coelagt Systes 14eets Limitattees Ass 1tsable for Heate lates e to 40*Me for the Servies Perted e to il IFFT SALas - Utrtt 1 3/4 4-34 knendment No. ';g , Initial RT4T * *II'I Ri g After il (F7Y: 1/4T
- 221 l'F 3/4T = 162'?
2290 2000 .F90 { t900 'iuG o ,,0 I- m u,fs ~, SOS g-p eu.i. 290 , s i a 300 feettafE3 feuptRATU E (SBS.F) 8 CINfAlls 18 l$nas!N FCR Pols!BLE Ill5falamit (RAGR$ neune 2.4.s Sale. Unit 1 Reestar Caelant System Cae10eun Liettations , Ase11 cable for Caeldema Rates e to 100*Me for the service Period g to 15 EFPT sasa . verst 1 3/4 4-27 Amendment No. 105 e I i . i .. . l . , *? TRIAL PROPERTV BASIS i t CONTROLLIWG MTERIAL: . LONGITUDINAL WELD l COPPER CONTENT: 0.35 WTt { NICKEL CONTENT: 1.00 WTs { INIT!AL ATNOT: 56'F ! RT NOT AFTER 10 EFPY: 1/4 T, 178.6'F ! 3/4 T. 116.1'F , e I l_ CURVES APPLICA8LE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERICO l UP TO 10 EFPY AND CONTAINS NO MARGIN FOR POS$18',E INSTRUE NT ERA M 5 i I i 2W_...... ...... ............................ ,.........., ,
- .e.........
til e. ...................... ........................ i r. e........... - . .....I...., .C.. .. ............................ ............ .. ..... i 280 s .. ,g .................................... ............ i 1184................................................. . .......................... l e C .......... ........e................ .e.4..... i w 12SI' i .........&. t......... ................................g..... .......... ........... ..............m........... ......... ......... r- ........................ ........................ .............. ... .e ' 4. 1 i W I 1884a .................................... ..................r.. 2. a.... = ee.e... > 3 9- .r ...e .................... 3 750 3 3 *. . .o.. ... ............. e.. l 3 ..ede...o..................... .. ..............e
- I 3 .................W........ ... .............
3 ............-.............. .. i 100 3. .i l -i 3 .............. -. . . . . . . . . ...= 2103............... -i
- -i...................
i .= .= . .= = ===.. === = j i .....h........................ 0i ...........t 4......................................s.......e.........=...... i .................a e4 ( leestates 7tw enates (ese.r) 9 I f Sale Unit 2 reactor coolant system heatup limitations applicable i.: for the first 10 EFPY with maximum heatup rate of 60*F/hr ! FIccar 3.4-2 i 4 SAI.EM UNIT 2 I 3/4 4-23
- Amendment No. 86 9
s i u - , . . .-. __. _ _-- . _ - - o - , MA?tal AL PRODER?v BA515 CONTROLLING MTERIAL: LOMITUO!NAL dELO COPPER CONTENT: 0.35 his MICKEL COMTENT: 1.00 Wit INITIAL RTNOT: - 56'F RT AFTER 10 EFPY: NOT 1/47. 178.6'F 3/4T. 116.1'F CURVES APPLICA4LE FOR COOLDOWN RATES UP TO 100'F/nt FOR THE SERY1CE PER100 UP TO 10 EFPY AND CONTAINS N0 MAGIN FOR P055tti.E INSTRUMEN 2$00 1710
- 2000 1750 7 1100 E
1210 IiO90 W i - 'IY.. ni g;. 300 g . 40 210 0, g g g a ge fieltattS T8tPCEat W (Ott.F) 8 lr Sale Unit 2 reactor coolant systs cooldown limitations apolicable for the first 10 EFPY FIGURr 3.4 3 sA:.tx . Uw:? 2 3/4 4-29 Amendment No. 86 REACTOR C001. ANT SYSTDt BASES Finally, the new 10CTR50 rule which addresses the metal temperature of the closure head flange regions is considered. This 10CFR50 rule states that the metal tes.peratu5*F for normal operation by at least 120 when20the ure exceeds press *fth'*l***1*"8 percent the preservice hydrostatic test pressure (621 pegg for Salum). Table B3/4.4 1 indicates that the limiting RT of 26 F occurs in the closure head flange,F is 148 at pressures greater than 621 psigofSalenUnit1.andtheminimusT1o These limits do not affect Figures 3.4-2 and 3.4-3. Although the pressuriser operates in temperature ranges above those for which there is reason for concern of non ductile f ailure.* operating limits are provided to assure compatibility of operation with the fatigue analysis 4' 17. -p. w$hurAstu Cbja :;uir gas. The OPERA 3II.ITY of two POPSs or an RCS vant opening of greater than 3.14 square inches ensures that the RCS will be protectd from pressure transients whichcouldexceedthelimitsofAppendixGto10gTRPart50whenoneormore of the RCS cold legs are less than or equal to 312 F. Either_ POPS has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F l above the RCS cold les toeperatures, or (2) the start of a safety injection ) pump and its injection into a water solid RCS. / SA1.Di - UNIT 1 B 3/4 4-11 Amendment No.108 ' d0 j1 FGRM NC.NA.AP.ZZ-0059 3 10CFR50.59 SAFETY EVALUATION Page 1 of 7 Revision 0 I.D. Numbers / Reference / Revision: TECH SPEC BASES 3.4.10.3 (UNIT 2)
Title:
POPS - MASS ADDITION OF AN ECCS PUMP Aeolicability: Salem l' Salem 3 (Gas Turbine) X Salem 2 . Hope Creek Common to Salem I & 2 Common to Hope Creek & Salem COMPLETION AND APPROVAL Preparer: - U- Date / f
~~
Peer Reviewer: y..-- Date 8/ . Approval: _ [4 Date /.2M<Nf SORC Review: 1#_ An Mtg No.bb Date /N/!fk G.M. Approval: +- Date / 7 U '
'/
1.0 10CFR50.59 REVIEW - 10CFR5 applies because: 1.1 The proposal changes the facility as described in the SAR. YES X NO Section 7.6.3.3 POPS Design Evaluation - Tech Spec. Bases 3.4.10.3 (Unit 2). 1 1.2 The proposal changes prc:cdures as described in the SAR. YES NO = X No procedures affected. 1.3 The proposal involves a test or experiment not described in the SAR. YES NO X Not a test or experiment. j Nuclear Common Rev.3
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j__ l ' _ a .__ J
Y FORM NC.NAJAP.ZZ-0059 3 10CFR50J9 SAFETY EVALUATION Page 2 of ' 7 Revision 0
!.D. Numbers / Reference / Revision: ._IFCH SPEC B ASES ' 3.4.10.3 (UNIT 2)
Title:
POPS - MASS ADDITION OF AN ECCS PUMP 2.0 TECHNICAL SPECIFICATION REVISION DETERMINATION - Does the proposal require a Technical Specification change? YES NO._X .__ If a change is required, STOP. Contact Nuclear Licensing for assistance in preparation of a License Change Request. l Identify the pertinent Technical Specification sections that were reviewed to make the determination: 3.4.10.3 Overpressure Protection System (Unit 2) 3.5.3 ECCS Subsystems Tavg < 350'F j
3.0 DESCRIPTION
3.1 Describe the modification or activity being evaluated r.nd its expected effects. The current Bases for Tech. Spec. 3.4.10.3 (Unit 2) and UFSAR Section 7.6.3.3 states that either POPS has adequate relieving capability to protect the RCS from overpreesurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam-generator less than or equal to 50'F above the RCS cold leg temperatures, or (2) the start of a safety injection pump and its injection into a water solid RCS. Because of re analysis associated with mass addition transients (performedga result of Westinghouse letter 3/15/93 - Ref. LER 94-017), I
- the Bases of thMech. Speci. ATeing changed to reflect a high head safety injection pump {
combined with a positive displacement (PD) charging pump. This change reflects actual mass addition transient assumptions based on actual plant operating procedures and administrative controls for preventing the injection of an intermediate head safety injection pump into a water solid RCS.
- This change is also being made to UFSAR Section 7.6.3.3 for POPS Design Evaluation. The net effect of this change is minimal since the flow of the combined high head SI pump and PDP is equivalent to an intermediate head SI pump. The P/r limits contained in TS will continue to be m et.
s 3.2 Identify the parameters and systems affected by the change.
- RCS peak pressure during a low temperature overpressure transient. (POPS)
Nuclear Common Rev.3 k J
y FORM NC.NA.AP.ZZ-0059 3 10CFR5039 SAFETY EVALUATION - Page 3: ef 7 ,
+
Revision 0
,I.D. Numbers / Reference / Revision: ' TECH SPEC BASES 3.4.10.3 (UNIT 2F
Title:
. POPS - MASS ADDITION OF AN ECCS PUMP ? U 3.3 Identify the credible failure modes associatedevith the change. _ Single failure of a POPS valve to open has been evaluated which is the most limiting
& single failure.
3.4 - Provide references to location of information used for the safety evaluat[$n. UFSAR Section 7, Technical Specifications, Unit i License Amendment #24 (including SER).
-3.5 Other Discussion, if applicable.
The POPS rystem Technical Specifications and Bases for Salem Unit I was -- implemented via license amendment #24L The Safety Evaluation Report was y
- submitted to PSE&G ty the _NRC on February 2,1980.1.nis Safety Evaluation stated that the most limiting mass input transient identified was inadvertent injection by the largest safety injection pump in a cold shutdown condition. A final pressure of 446-psig was postulated to occur for this worst case mass input transient. A number of provisions for prevention of pressure transients were placed in Salem's operating -
g procedures to limit low temperature overpressurization such as: Pressurizer bubble for stanting RCPs, taw...g out of both intermediate head safety irQection pumps when RCS - temperature is below 350'F or closing the SI pump discharge' isolation valve to prevent injection, tagging of one high head SI pump, and providing ventpaths during intermediate head safety injection pump testing! Salem Unit 2 was licensed with POPS ins +slied, ne Brsses for Salem Unit 2 is similar to Salem Unit 1. Due to a Westinghouse NSAL which addressed non conservatisms in the calculated p peak pressure for Salem, an additional 31 psig was added, u An additional administrative limitation of 1 RCP inservice below 200*F _was imposed due to the NSAL. In re-analyzing the non-conservatisms, PSE&G used the GOTHIC computer code in place of the Westinghouse algorithyms. Revised actual Salem pump flow data was used as inputs into this code rather than the values originally used.- The resulting analyses demonstrated acceptable peak pressures for injection of either an intermediate head safety injection pump alone or injection of a high head safety injection pump combined with a positive displacement pump for Unit 2. Nuctear Common Rev.3
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, FORM NC.NA AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 4 of 7 Revision 0 I.D. Numbers / Reference / Revision: TECH SPEC BASES 3.4.10.3 (UNIT 2)
Title:
EQPS - MASS ADDITION OF AN ECCS PUMP 4.0 USO DETERMINATION - !s an Unreviewed Safety Question (USQ) involved? , 4.1 Which anticipated operational transients or postulated design basis accidents previously evaluated in the SAR are considered applicable to the proposal? Reactor vessel peak pressure at low temperature conditions. L 4.2 May the proposal:
- a. Increase the probabi:ity of an accident previously evaluated in the SAR?
l YES NO X DISCUSSION: i. The probability of any accident previously evaluated in the UFSAR will not incre'ase because no plant physical changes are being made or no system performance related changes are being made.
- b. Increase the consequences of an accident previously evaluated in the SAR?
YES NO X DISCUSSION: The Bases of Tech. Spec. 3.4.10.3 (Unit 2) are being revised to reflect the actual limiting mass addition transient assumptions from the injection of a high head saf,ty injection pump in conjunction with the running positive displacement pump. These assumptions are based on actual operating procedure requirements. The consequences of this mass addition transient, based on the current operating procedures and administrative controls, is appropriate. 'Ihe consequences of this event are not changed since the P/r limits provided in Tech Spec 3.4.10.1 (Unit 2) are maintained. Nuclear Common , Rev.3
;o 7 FORM NC.NAiAP.ZZ-0059 3 .
10CFR$039 SAFETY EVALUATION = Page 5 ef 7 Revision 0
' !.D.i Numbers / Reference / Revision: TECH SPEC BASES 3.4.10.3 (UNIT 2)
Title:
POPS - MASS ADDmON OF AN ECCS PUK(P
- 4.3 - What malfunctions of equipment important to safety that were previously evaluated in the --
SAR are considered applicable to the proposal?,
. DISCUSSIONi Only the malfunction of the SI system L
isolation) and the malfunctionurizer Ia press (which.may~ PORV and initiate a POPS actuation du its associated instrumentation (which mitigates the pressure increase) are applicable to this proposal.-
- 4.4 May the proposal:
'a. Increase the probability of occurrence of a malfunction of equipment imoortant to safety previously evaluated in the SAR?
YES -NO X -l DISCUSSION:
.The probability of the malfunction of a PORV is not increased due to the changes to the
-Bases of the flow rates.- The systems are not being physically modified and therefore will not -
be_ impacted. The failure assumptions of the original analyses remain valid Therefore, this
- change will not increase the probability of occurrence of a malfunction of equipment -
important to safety previously evaluated in the SAR. b.' Incrsase the consequences of malfunction of equipment imoortant to safety previously evaluated in the SAR?. 4 YES NO X DISCUSSION: I Tha consequences of a malfunction of equipment important to safety is not increased. The : failure assumptions used in the original POPS ' analysis remain unchanged. The SI system is not being physically modified and therefore will not be impacted. The PORVs and their instrumentation are also not being modified and therefore are not impacted.
' Noelear Commen Rev.3 e
a
/
' FORM NC.NA AP.ZZ 0059 3 13CFR50.59 SAFETY EVALUATION Page 6 of 7 Revision 0
!.D. Numbers / Reference / Revision: TECH SPEC BASES 3.4.10.3 (UNIT 2)
Title:
POPS - MASS ADDITION OF AN ECCS PUMP 4.5 Max the proposal:
- a. Create the possibility of an accident of a different tvoe from any previously evaluated in the SAR?
YES NO X % DISCUSSION: Operating procedures required that SI pump power be removed in Mode 4 and 5. This change reflects actual mass addition transient assumption based on plant operating pncedures. This is not an accident of a different type from that proposed in the SAR, but the same mass addition transient based on actual operating conditions. In instances when intermediate head safety injection pump availability is necessary (e.g. flow balance testmg and SW header outage at mid loop), an additional vent path is required by procedure to preclude overpressurization, or the pump discharge isolation valve is maintained in the closed position.
- b. Create the possibility of a malfunction of a different tvoe from any previously evaluated in the SAR?
YES NO X DISCUSSION: The possibility of a malfunction of a different kind is not possible, since no changes are being made to plant systems. The only difference is the assumptions for the mass addition. These are based on actual plant operating procedures. Herefore, this will have no effect on system malfunctions. Nuclear Common Rev.3
j FORM NC.NA.AP.ZZ 0039 3 10CFR50.59 SAFETY EVALUATION Page 7 of 7 Revision 0 1.D. Numbers / Reference / Revision: TECII SPEC BASES 3.4.10.3 (UNIT 2)
Title:
POPS - MASS ADDITION OF AN ECCS PUMP ,, 3 4.6 Dssa the proposal reduce the margin of safety as defined in the basis for any Technical Specifications? YES NO X Discuss the bases for the determinations and identify the pertinent Technical Specification sections that were reviewed to make the determir.ation (use continuation sheets if required). The margin of safety is not reduced. The change to the mass addition transient assumptions reflects the actual plant and operating procedure requirements (i.e., those pumps that might be available in Modes 4 and 5). The P/r limits are not exceeded. The net effect of this change does not reduce the margin of safety since the flow of the combined high head SI pump and PDP is equivalent to an intermediate head SI pump.
5.0 CONCLUSION
If ALL answers in Section 4 are "EQ" the proposal does HQIinvolve a USQ. If ANY answer in Section 4 is "YF1" the proposal involves a USQ. Is a USQ involved? YES NO X If a USQ is involved, obtain assistance from Licensing for additional processing. LCR Number:
7, FORM NC.NA AP.ZZ-0059 3 10CFR50.59 SAFETY EVALUATION Page 1 of 7 Revision 0 1.D. Numbers / Reference / Revision: TECH SPEC BASES 3.4 9,3 (UNIT 1)
Title:
POPS MASS ADDITION OF AN ECCS PUMP Anolicabilily;;_ X__ Salem i Salem 3 (Gas Turbine) Salem 2 liope Creek Common to Salem 1 & 2 Common to Hope Creek & Salem COMPLETION AND APPROVAL Preparer: l8cWAffg~ Date t/&/77 ( / Peer Reviewer: ze#mt f7 Date d M 5T Approval . Date 2I[JJ SORC Review: A\ 31k Mtg No.75%d Date z l7 h# G.M. Approval: , l h,w3 Date 2!dS[ Pert. 9 e 1.0 10CFR50.59 REVIEW W).10CFR50.59 applies because:
\
1.1 The proposal changes the facility as described in the SAR. YES X NO Section 7.6.3.3 POPS Design Evaluation - Tech Spec Bases 3.4.9.3(Unit 1). 1.2 The proposal changes procedures as described in the SAR. YES NO X No procedures affected. 1.3 The proposal involves a test or experiment 'not described in the SAR. YES NO X gQ Not a test or experiment. j i Nuclear Common Rev.3
?
l 8 . _ _ _ .___. _. .. . l
FORM NC.NA.AP.ZZ 0059 3 10CFR50,59 SAFFlrY EVALUATION Page 3 of 7 Revision 0 1.D. Numbers / Reference / Revision: TECH SPEC BASES 3 4.9.3 (UNIT 1)
Title:
PQ.fj - MASS ADDITION OF AN ECCS PUMP 3.3 Identify the credible failure modes associated with the change. Single failure of a POPS valve to open has been evaluated which is the mest limiting single failure. The mass addition transient could be initiated due to a malfunction of the Si system. 3.4 Provide references to location of information used for the safety evaluation. UFSAR Section 7 Technical Specifications, Unit 1 License Amendment #24 (including SER), 2ech. Spec. Bases Change 3.4.10.3 (Unit 2). L 3.5 Other Discussion, if applicable, , The POPS system Technical Specifications and Bases for Salem Unit I was implemented [ via license amendment #24. The Safety Evaluation Report was submitted to PSE&G by . [ the NRC on February 2,1980. This Safety Evaluation suted that the most limiting mass input transient identified was inadvertent injection by the largest safety injection pump in a cold shutdown condition. A final pressure of 446_ psig was calculated to occur for this worst case mass input transient based on Westinghouse algorithm. A number of provisions for prevention of pressure transients were placed in Salem's operating procedures to limit low temperature overpressurization such as: Pressurizer bubble for starting RCPs, tagging out of both intermediate head safety injection pumps when RCS temperature is below 350*For closing the SI pump discharge isolation valve to prevent injection, tagging of one high head SI pump, and providing vent paths during intermediate %i sfety injection pump testing. Salem Unit 2 was licensed with POPS installed.. Due to a Westinghouse NSAL v1ich addressed non-conservatisms in the calculated peak pressure for Salem, an additional 31 psig was added based on the operation of one RCP. This resulted in an additional administrative limitation of 1 RCP in service below 200*F. Subsequently, it was determined that if the PD pump was in service when an inadvettent Si occurred, the PD pump would continue to operate resulting in the exceedance of the P/T limits with RCS temperature < 200 degrees F. This was reported to the NRC under LER 272/94-017. In re.-analyzing the non-conservatisms, PSE&G used the GOTH!C computer code in place of the Westinghouse algorithm. Revised actual high head SI pump flow data was used as inputs into this code rather than the values originally used. The resulting analyses demonstrated acceptable peak pressure for Nuclear Conunon Rev.3 j
- p. FORM NC.NA AP.'tZ-0059-3 10CFR56.59 SAFETY EVALUATION Psge 4 of 7 Revision 0 1.D. Numbers / Reference / Revision: TECF SPEC BASES 3.4.93 (UNil 1)
Title:
POPS - MASS ADDITION OF AN FCCS PUMP _ injection of a high head safety injection pump. Admin. Controls ensure that only the injection of a high head Si pump will be the most limiting mass addition transient for Unit 1. 4.0 USO DETERMINATION - Is an Unreviewed Safety Question (US) involved? 4.1 Which anticipated operational transients or postulated design basis accidents previously evaluated in the SAR are considered applicable to the proposal? Reactor vessel peak pressure at low temperature conditions. 4.2 hiaL the proposal:
- a. Increase the probability of an accident previously evaluated in the SAR?
YES NO X DISCUSSION: The probability of any accident previously evaluated in the UFSAR will not increase because no plant physical changes are being made or no system performance related changes are being made.
- b. Increase the consequences of an accident previously evaluated in the SAR?
YES NO X DISCUSSION: The Bases of Tech. Spec. 3.4.9.3 (Unit 1) is being revised to reflect the actual lintiting mass addition transient assumption from the injection of a high head safety injection pump. This change is similar to a Bases change for Unit 2 with the exception that a running positive displacernent pump is not assumed. The positive displacement pump
- will be administratively controlled to ensure that power is removed wnen operating in Mode 5 (< 200 degrees F) and a vent path has not Leen established per Tech. Spec ,
3.4.9.3. The consequences of this mass addition tranuent, eased on the operating procedures and administrative controls, is appropriate. The consequences of this event are not changed since the P/T limits provided in Tech Spec 3.4.9.1(Unit 1)are maintained. Nuclear Common Rev.3 I i
-) FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 S AFETY EVALUATION Page 5 of 7 Revision 0 l.D Numbers / Reference / Revision: TECH SPEC BASES 3.4 9.3 (UNIT D
Title:
POPS MASS ADDITION OF AN ECCS PUMP 4,3 What malfunctions of equipment important to safety that were previously evaluated in the SAR are considered applicable to the proposal? DISCUSSION: Only the malfunction of the Si system (which may initiate a POPS actuation due to letdown isolaticn) and the malfunction of a pressurizer PORV and its associated instrumentation (which mitigates the pressure increase) are applical!e to this proposal. 4.4 May the proposal:
- a. increase the probability of occurrence of a malfunction of equipment important to safety presiously evaluated in the SAR?
YES NO .< DISCUSSION: The probability of the malfunction of a PORV is not increased due to the changes to the Baws of the 'ma-. ddition flow rate. Th: systems are not being physically modified and therefore will not be impacted. The failure assumptions of the original analyses remain valid. Therefore. this change will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.
- b. Increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR?
YES NO X_ DISCUSSION: The consequences of a malfunction of equipment important to safety is not increased. The failure assumptions used in the original POPS analysis remain unchanged. The Si system is not being physically modified and therefore will not be impacted. The PORVs and their instrumentation are also not being modified and therefore are not impacted. Nuclect Common Rev.3 )
l # FORM NC.NA AP.Z7-0059 3 10CFRSO.59 SAFETY EVALUATION l Page 6 of 7 Revision 0 l I.D. Numbers / Reference! Revision: TECH SPEC BASES 3.4.9.3 (UNIT 1)
Title:
POPS - MASS ADDITION OF AN ECCS PUMP 4.5 hhy.the proposal:
- a. Create the possibility of an accident of a different tvoe from any previously evaluated in the SAR?
YES NO X DISCUSSION: Operating procedures and Admin. Controls will ensure that the intermediate head SI pumps. one high head SI pump, and the PD pump have power removed in Mode 5. This change reflects the actual mass addition transient assumption based on plant operating procedures. This is not an accident of a differer.t type from that propose 3 in the SAR, but the same mass addition transient based on actual operating conditions. In instances when intermediate head safety injection pump availability is riecessary (e.g. flow balance testing and SW header outage at mid loop), an additional vent path is required by procedure to preclude overpressurization, or the pump discharge isolation valve is maintained in the closed position to ensure that the mass addition transient assumption continues to be met.
- b. Create the possibility of a malfunction of a different tvoe from any p eviously evaluated in the SAR?
YES NO X DISCUSSION: The possibility of a malfunction of a different kind is not possible, since no changes are being made to plant systems. The only difference is the assumptions for the mass addition. These are based on actual plant operating procedures. Therefore, this will have no effect on system malfunctions. Nuclear Common Rev.3 y _ _ - - - - - - - _
j~ FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUAT' ION Page 7 of 7 Revision 0 I.D. Numbers / Reference / Revision: TECH SPEC BASES 3.4 9.3 (UNIT 1) _. Titler POPS - MASS ADDITION OF AN ECCS PUMP 4.6 Does the proposal reduce the margin of safety as defined in the basis for any Technical Specifications? YES NO X 1 Discuss the bases for the determinations and identify the pertinent Technical Specification sections that were reviewed to make the determination (use continuation sheets if required). The- margin of safety is not reduced. The change to the most limiting mass' addition transient assumption (the injection of a high head SI pump into a water solid.RCS )- based on plant operating procedure requirements and administrative - controls ensures the- P/T limits are not exceeded. The injection of a high head SI pump into a water solid RCS based on realistic pump flow rates results in a peak pressure that is less than-that originally calculated using the Wes:inghouse algorithm. Therefore, the' net affect of this change does not reduce .the margin of safety.
5.0 CONCLUSION
l If ALL answers in Section 4 are "NO."the proposal does NOT involve a USQ. 4 If ANY answer in Section 4 is "YES "the proposal involves a USQ. Is a USQ involved?
-YES NO X If a USQ = is involved, obtain assistance from Licensing for additional processing.
LCR Number: Rev. 3 Nuclear Common
O i REACTOR COOlJLW SYSTD4 EMES Finally, the new 10CFR50 rule which addresses the metal temperature of the closure head flange regions is considered. This 10CFR50 rule states that the metal temperatuge of the closure flange regions must escoed the material RT byatleast120Ffornormaloperationwhenthepressureexceeds20 percent $ the preservice hydrostatic test pressure (621 pegg for Sales). Table B3/4.4 1 indicates that the limiting RT of 28 F occurs in the closure head flange,ofSalemUnit1,andtheminimumT1owabletemperatureofthisregion is 148 F at pressures greater than 621 psig. These limits do not affect Figures 3.4-2 and 3.4-3. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure,* operating limits are provided to assure compatibility of operation with the fatipus analysis performed in accordance with the ASME Code requirements. The OPERABILITY of two POPSs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protectd from pressure transients whichcouldexceedthelimitsofAppendixGto10gTRPart50whenoneormore of the RCS cold legs are less than or ecuel to 312 F. Either POPS has adequats relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam gaaerator less than or equal to 50*F above the RCS cold leg temperatures, or (2) the start of a" safety injection pump and its injection into a water solid RCS. l H.3h tid l l l SALDi - UNIT 1 B 3/4 4-11 fcendment No.108
- t. L
;.c A; testing provision in the POPS circuitry allows - for test opening- .
- of Jehe relief valves prior .to use of .the . system below 3120F, The
" TEST * ' pushbutton. . when depressed. will operate the - rellef valve' i
provided that the associated taotor operated . ' valve is : closed,
~
Other portions- of - the POPS can.be tested in' a- manner similar to
-3' other protection system functions; The existing power operated ; relief valves (PORVs) are utilized for-overpressure protection at' low temperature in Units 1 and 2.
7.6.3.3- Desien Evaluation > The POPS is designed as a " protection grade" system in accordance f with the applicable portions of -IEEE Standard 279 1971. The us'e of proven devices provides assurance that the system is compatible with other Protection System . equipment; The use of administrative controls to arm the -POPS -is considered acceptable due to the expected infrequent need for' overpressure protection - at_ low temperature. The ' POPS relief valves protect the ' RCS from pressure t rat.41ents -- which could exceed the limits of'Appendin G to.10CFR50 when one'or - more.RCS cold leg temperature is at or below 3120F. ' Ei'.he r _ POPS has adequate relieving. -capacity to protect thr. RCS from-overpressurization as a - resule of the n limiting 1. eat ' input or mass input cases: (1) the . start of an idle Rhetor Coolant Pump with: the secondary water - temperature of - the- steam generator - less ' than or equal to 500F above RCS cold _ leg temperature or (2) the start of tif Heel a safety _ injection - pump and its injection into_ the water solid RCS 4. AL number of provisions for prevention of pressure transients below 'e
! TNMW P (when the ' RCS temperature is.below 3120F) presently exists _ in j the Technical Specifications.
t "1" '
. AHad4).
In order to cause an unwanted relief - valve opening at normal
< operating pressures, an operator would'have to erroneously arm the POPS system. This would require bypassing the administrative control of the key associated with the keylocked pushbutton 7.6-6 SGS UFSAR Revision 6
' February 15, 1987 j'
o L 1
e INSERT 1 TO UFSAll SECTION 7.6.3 3 (Unit 1 only), or the start of an Intermediate llead Safety injection pump and its injection into a water solid RCS (Unit 2 only), or the start of a High Head Safety injection pump in conjunction with a mnning Positive Displacement and' injection into a water solid RCS (Unit 2 only). (Note: - The Unit 2 Changes to the POPS mass addition assumptions were previously af, roved by SORC on 12/21/94, and are reflected here for consistency. Only the Unit I changes are a addressed by this Safety Evaluation.) l l 4 I
J( , TO: D. Smith Principal Engineer - Nuclear Licensing FROM: K. O'Gara Licensing Engineer
SUBJECT:
EVALUATION OF REPORTING REQUIREMENTS AND ROOT CAUSE ASSESSMENT ASSOCIATED WITH PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS) ISSUES (IR 95-343) DATE: April 6, 1995 The Offsite Safety Review (OSR) Group initiate'. Incident Reports 95-343 dated 04/04/95 and 95-398 dated 04/13/95. IR 9S-348 documents two (2) instances where PSE&G failed to report to the NRC in accordance with 10CFR50.72/73 that the Salem POPS was outside of its design basis. The two instances are related to evaluations performed by Nuclear Mechanical Engineering (NME) to address the POPS setpoint non-conservatisms that were identified by Westinghouse in NSAL 93-005B dated March 15, 1995. IR 95-398 documents four occasions where PSE&G failed to perform a 10CFR50.59 Safety evaluation when the original design basis of the POPS was changed to address the setpoint non--conservatisms. The purpose of this memo is to document Licensing's position with regard to the reportability of the 2 instances cited by OSR in IR 95-343, and to determine if 10CFR50.59 Safety Evaluations were warranted as identified in IR 95-39C. This memo also.provides a discussion of the root causes that'may have contributed to these deficiencies occurring. IR 95-348 Westinghouse identified a concern that the pressure differential associated with the operation of RCPIs) during the mass addition transient was not considered. The initial POPS ar.alysis for Salem Units 1 and 2 did not account for the pressure differential between the mid-plane of the core and the location of the pressure sensors 1;cated at the Reactor Coolant System hot legs with the RCPs in operation. To quantify the effect_ on Salem, specific pressure differences associated with RCP operation were
. calculated for 1, 2, and 4 RCPs operating. The results of these calculations provided pressurc differences of 31, 39, and 73 psig for 1, 2, and 4 RCPs, respectively. These values include a 2.0 psig correction for transmitter elevation differences.
Evaluation of the Westinghouse results was documented in a memo from NME on 12/30/93. C
\)
i
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i
,f 2
( OSR concluded in IR 95-343 that on 12/30/93, Salem Units 1 and 2 I were_outside of the design bases for POPS when considering the l pressure differential associated with the operation of the RCPs. l The allowable peak pressure limits determined in accordance with 10CFR50, Appendix G for Salem Units 1 and 2 are 450 and 475 psig, respectively. These limits are reflected in the Pressure-Temperature curves contained in T/S Figures 3.4-2 and 3.4-3 for Units 1 and 2, respectively. The calculated peak pressure for both units for the mass addition transient was 446 psig at that time. If you add the pressure oifferential associated with the operation of a minimum of one RCP, the peak pressure would be 477 psig. T eve, both plant's pressure limits would be exceeded. Therefore, concluded that on 12/30/93 the design basis for POPS could no be met based on the Westinghouse analysis results, and this shoul have been reported to the NRC in accordange with 10CFR50.72/73. NmE u/u/ Cat,. L e.r/v d Ob. a hA b & A==64m- P daNi: - "
# 4 cd W A
- J'fg e nC4 6 M] MMC On May 26, 1994, NME completed fu"r'th'$I~ evaluation of the POPS issue. The re ults of this evaluation concluded that the pressure differential assoc.ated with the RCPs in operation did not need to be considered since the RCPs are not operated with the pressurizer in a water solid condition. The results of the i NME evaluation determined that for the mass addition transient, i
the Unit 1 peak pressure would be 450.7 psig. This exceeds the pressu re limit for Salem Unit 1 (450 psig) by 0.7 psig. , Therefore, OSR concluded that at this time Salem Unit 1 POPS was outside its design basis, and this should have been reported to the NRC in accordance with 10CFR50.72/73. To support Licensing's evaluation of the reportability of the two instances cited by OSR in IR 95-343, the following additional information is provided: On September 27, 1994, NME documented further evaluation of the POPS nonconservatism issue and evaluated the mass addition transient assuming that the RCPs could be in operation prior to an inadvertent SI actuation resulting in the pressurizer going water solid. This evaluation was performed using a flow rate associated with the CCP since both IHSI pumps would have power removed. On November 17, 1994, PSE&G did notify the NRC that the Salem Unit J POPS was considered outside of its design basis. Thio sas due to a new analysis assumption that the PDP may be running during a mass addition transient that resulted from an inadvertent SI actuation. Under this new analysis assumption, the Unit pressure limit of 450 psig would be excerted. This deficiency was reported to the NRC under LER 94-17. The Unit 2 pressure limit was still maintained.
't 3
The NRC issued a Safety Evaluation Report (SER) for Salem Unit 1 to add POPS to the Technical Specifications on 02/21/80. This SER documents assumptions that were utilized in the POPS analysis for Unit 1. Although the SER is not applicable to Unit 2, similar analysis assumptions were used in support of the Unit 2 operating license. The NRC SER discusses the administrative controls in place to remove power to the SI pumps to limit the the affects of the mass addition transient. However, the mass 1 addition transient analysis was completed assuming the worst case l condition that the higher IHSI pump flow would be injecting into . a water solid pressurizer. Also, the SER discusses operation of ) the RCPs with a bubble in the pressurizer. This was , administratively controlled to limit the affects of the mass addition transient. However, the mass addition transient analysis was completed assuming the worst case condition that the pressurizer was water solid during this event. This assumption is contained in the current T/S Bases. IR 95-343 was supplemented by memo OSR 95-012 dated April 10, 1995. This memo documents that PSE&G was aware that that POPS was outside of its design basis as early as September 29, 1993 by a Letter from Westinghouse. The IR states that reportability should have been addresses by PSE&G at this time. The analysis results provided by Westinghouso formed the basis for the memo generated by NME on 12/30/93. Generic Letter 91-18 discusses the use of 10CFR50.59 Safety Evaluations to address degraded or Non-conforming conditions. GL 91-18 states that the 50.59 process may be used as part of the Corrective Action Program in lieu ci rectoring the affected equipment to its original design. The 50 53 process may be used as long as such a change does not require a License Amendment or result in an Unreviewed Safety Question. GL 91-18 does not relieve the Licensee of the responsibility to report the original condition in accordance with 10CFR50.72/73 or other reporting mechanisms. A copy of the applicable Section of GL 91-18 is attached for your information. (It is noted taat further discussion of the completeion of 10CFR50.59 Safety Evaluations to address this issue is conteined under IR 95-398.) Based on the above discussion and review of the NRC Safety Evaluation Report For Salem Unit 1 dated 02/21/80 Ohat added POPS to the Unit 1 Tech. Specs., it should be concluded that both Salem Units were outside of their design bases during the period of time between September, 1993 and the dates that the T/S bases was changed for Uait 2 (12/21/94) and for Unit 1 (2/06/95) for purposes of reporting in accordance with 10CFR50.72/73. This conclusion is consistent with GL 91-38.
- r w
l i 4 l The notification to the NRC for Unit 1 being'outside the design ! bases only covers'the period of_ time between November 17, 1994 ! and the date that the Bases was changed. Also, this notification i . only addresses the issue regarding operation Of the PDP. IR 95-398 , I Between-the time that NME completed the initial evaluation of the : POPS non-conservatism issue in response to Westinghouse NSAL 93- l 005B on 12/30/93,-and the time that PSE&G documented the changes - to the POPS Tech. Spec. Bases for Unit 2 or reported the PDP ; , deficiency for Unit 1, the following analysis assumptions : i contained in-the SER were revised without the completion of a i
- 10CFR50.59 Safety Evaluation as documeneted in IR 95-398:
1 The NME memo of 12/30/93 discussed administrative limits on .
- the maximum number of RcPs that could be inservice when the !
RCS tempertaure was below 200*F. This changed the design
' basis as contained in the NRC SER dated 2/21/80. i The May 26,-1994-memo from NME was based on a Computer Code .
(GOTHIC) which was not reviewed and approved-by-the NRC in : . the SER. The SER conclusicns were based on the LOFTRAN code ! 4 provided by Westinghouse. -Also, the number of RCPs was ; again reduced ta a maximum of one without completeing a 50.59 SE.- ,
- i. ?
- On September 27, 1994, theilower flow rate of the
- Centrifugal Charging Pump (560-gpm) was used to re-analyze '
the peak pressure from the mass addition transient. This differed from the NRC accepted design flow of 780 gpm as discussed in the NRC SER. -
- On: November-17, 1994, the POPS design transient was again revised to address Positive Displacement pump (PDP) operation which was also not considered-in the NRC SER.
E !~ Also,-it is noted that the May 26, 1994 memo-from NME did not 4 - consider the affects of RCP operation because the RCPs would not , ~ be started-with the pressurizer in a water solid condition. This Lis also different than-the assumptions contained in the NRC SER
. dated 2/21/8,0 although not included in the scope of IR 95-398. :
. s As discussed _above, Generic Letter 91-18 discusses the use of 10CFR50.59 Safety.. Evaluations to address degraded or Non-conforming conditions. GL 91-18 states that the 50.59 process , may be used as:part of the C0rrective Action Program in lieu'of
- restoring'the-affected equipment-to its original design. The I
4
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i. 5 50.59 process may be used as long as such a change does not require a License Amendment or result in an Unreviewed Safety Question. Therefore, it is agreed that any changes to the POPS analysis should have received a 10CFR50.59 SE to document the ' ch.nges to the POPS design basis as documented in the SER. This would include changes to the T/S basis which were completed in accordance with 10CFR50.59. ! A . , v. Csa. Hv m *h~* *l N L e v. rW Root Cause AgAggsment y, ( . ksome e 'p g , t w t.. p L
- Each of the issues identified in IR's 95-343 and 95-398 can be attributed to failure to follow procedures or the failure.,,etg r %,
current NRC regulations. Both of these failures are considered weaknesses in human performance. The factors that may have contributed to these weaknesses include:
-kdeggtg understanding of the POPS Design Basis Al W
- bdequateunderste.ndingofwhat is considered "outside of design basis" for reporting purposes ple"
- Understanding when an IR/DEF is required, or when an issue should e entered into an appropriate Corrective Action Program
- erstanding when a 10CFR50.59 SE should be prepared ,e' V i l+l~d ""
p, Properly prioritizing the review of issues which may have / salety significance or require further operability reviews (UL M ** e
*****)
,a w a.g sk. qW ,
x M 24 ?that ce[.ainASMECodeCases ire NRC approval prior to their use/ V^
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,_ ~ _. - .. _ _ . , - . _ . - _. _ . . _ . _ _ - .. _ , . . . _ _ . - . _ _ . _ . . . . _ _ . _ . _ . _ _ a e . ; .k , ATTACRMENT 1
SUMMARY
OF SYNERGY REPORT FINDINGS [ FINDING 1
~
2
-Use of thafGOTHIC computer code in-reanalyzing the POPS peak l
, pressure had not been reviewediand-approved for use by the NRC. .!
.The LOFTRAN computer code was acknowledged as acceptable in the '
SER issued by the NRC.- Therefore, use of-the GOTHIC Code could j be_ considered-outside the design.(licensing) basis of the. POPS t and a'10CFR50.59 Safety Evaluation should have been completed. .! FINDING 2 Failure by Engineering to initiate an. Incident Report documenting that the. POPS was outside its design (licensing) basis when Engineering received a letter from Westinghouse-(9/93) presenting the_ plant specific results for Salem. These results indicated that the existing P/T limits would be exceeded without changing , the current-POPS setpoint, , FINDING 3 [ Based on Finding 2, PSE&G should have reported this deficiency'to
- the NRC in'accordance with 10CFR50.72/73 for the plants being ,
outside the Design basis.- i
- FINDING 4
' Engineering disposition in December,' 1993 of the Westinghouse plant specific analysis results relied upon an ASME Code Case-
-which was not approved by the.NRC for use at Salem, j
~ FINDING 5-
- i v l Although not discussed in detail,' reference is made to a draft IR ,
i prepared by-Salem Technica1' documenting the design issue that-was !
. not: issued.' The may have addressed the Reportability-finding i
, discussed above. (It is noted that the failure to issue this IR =
may be considered a failure to follow procedure.)- FINDING 6 , r 1 Licensing was aware of the design deficiency.and should have
-initiated an'IR which would have documented a proper reporting i determination. (It is noted that a Licensing had drafted an IR l which was not issued.)'
P b 1
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- - - - . _ . - , - , . . ,- .,a.-~
r * , I ATTACHMENT 1 (Cont'd)
SUMMARY
OF SYNERGY REPORT FINDINGS EINDING 7 The May 26, 1994 Engineering memo allowed a " minor exceedance" of the P/T limits by 07 psig. This is outside the NRC approved licensing basis of the POPS. Changes to the Basis supported by the GOTHIC computer code should have been completed in accordance 10CFR50.59. (See Finding 1) FINDING 8 Changes to the flow rates used in the POPS analysis in 1977 should have been evaluated in accordance with 10CFR50.59. This includes both the realistic HHSI flows and the Centrifugal charging pump flows that were governed by T/S. This should have been considered a change to the licensing basis. FINDING 9 Since changes to the POPS design basis were necessary to continue to ensure the P/T-limits were not exceeded, operation of the plant could be considered outside the design basis and this should have been considered for reportability (i.e., since 9/94 when the realistic flow rates were used.
6-07-1995 3 42PM FR*.Ri PSEG OPER LICENS1tc 609 339 St.25 P 1 .'* p), PSE&G NUCLEAR LICENSING OPERATIONAL SUPPORT GROUP NUMBER OF PAGES (including cover sheet) T DATE ( S' TO: MFN O/6 W GROUP: /tcsm'.cfwc. PHONE / FAX NUMBER: /YV7 __ FROM: Md6" A /__ GROUP: PHONE NUMBER: COMMENTS C.d3s%Ao v M e Nd_ _ LAM ADAMUJ W- Ao h tm Y, m b Yf--- /Ls 0 - ah1 W << A _.
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August , 1995 U. S. Nuclear Regulatory Commission Document control Desk washington, DC 20S55 Attn.: Document Control Desk SALEN GENERATING STATION LICENSE Not DPR-70 DOCKET NC: 50-272 UNIT NO: 1 LICENSEE EVENT REPORT NO. 94-027-01 This Licenseo Event Report is being submitted pursuant to the requiraments of code of Federal Regulation 10CTR50. 73 (a) ( 2 ) (1) (B) . Sincerely, Clay C. Warren-General Manager - Salem operations KJPJ va SORC 95-~~- C Distribution LY.R Fila
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8-07-1995 3143P' FPC't P?iG CDER LICEN$lfC 609 339 St.D5 P. 3 e' l I I On 11/17/94, it was determincd that the following realistic assumptions could place Unit 1 outside.the design / licensing basin for Pressurizer overpressure Protection System (POPS) analysis should a safety injection (SI) signal occur Reactor Coolant System (RCS) temperature < 200*F, a Reactor coolant Pump (RCP) in operation, Positive Displacement Charging Pump (PDP) in service, and power available to a maximum of 1 Centrifugal Charging Pump (CCP) . Under these conditions, an SI signal could result in a peak RCS pressure of 474 psig which exceeds the current design basis pressure limit of 450 psig. This concern was not consitered applicable to Unit 2. Subsequently, it was determined that both Units 1 and a were outside their design / licensing basis for .the POPS analysis should a FI signal occur, as of December 30, 1993. / This conclusien is based on the differential pressure from the operating RCPs that w4s not considered in .the origina1JoPL analyripifor the mass addition transient. The adifflonal pressure from one or more RCPs operating (31 psig for i RCP) when added to thappe k pressure
-for this-transient (446 Psig) would exceed t P/T - mits of 450 psig and 475 psig for Units 1 and a respect oly./
6-07-19&B 3!45P" rR.H pr.tc opID LICENSitG 608 33D Gt.15 p. ca 6 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salen Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 2 of 5 Plant _and system Identificati_ gal Westinghouse - Pressurized Water Reactor Energy Industry Identification System (ETIS) codes are shown in the text as {xx) Identification of occurregggi Inadequate Margin Tor Pressurizer overpressure Protection During Low Temperature conditions (Applicab)eJo Uynitgd Event Date: November 17, 1994 ik M k k 4+ ' " "
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Report Date August, 1995 This report was initiated by Incident Report 94-419 Initial conditions: Mode 3 Reactor Power 100% Unit Load 1150 MWe D11911ption of occurrencer The current bases for Technical specifications (TSs) 3/4.4.9.3 (Unit 1) and 3/4.4.10.3 (Unit 2) states that one Pressurizer overpressure Protection $feten (POPS) relief valve, at a lift setting of </= 375 psig, provides adequate relieving capacity in the event of an-overpressure transient that includes inadvertent start of a safety injection (SI) pump (mass addition transient) into a water solid Reactor Coolant Sfsten (RCS). Subsequently, it has been determined that the following realistic nass addition transient assumptions could place b+ih wai^ outside the design t and licensing basis POPS analysis should an I signal occur: p( t - RCS temperature </= 200'P M i.
- One Reactor Coolant Pump (RCP) in operation
- Positive Displacement =-Charging Pump (PDP) in service
- Power supply available to a maximum of One Centrifugal Charging Pump (CCP)
6-0?-1995 1 d.2 " FROtt PSEG OMR LICENS!fG 607 330 5415 PB 6 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salen Generating Station Docket Number Lrt Wuaker Page Unit 1 50-272 , 94-017 3 of 5, Reserlotion of occurranear teont' d) At 1746 hours en November 17, 1994, the Nuclear Regulatory Commission was notified of this event, pursuant to the requiremente of 10CTR50.72 (b) (1) (11) . Under the above conditions, an SI signal could result in a combined flow fron the PDP and the CCP vith a 7eak RCS pressure of 474 psig. This exceeds the current des :ih pressure limit of 450 or Salem Unit 2. n#
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8 se gineering evaluation c sple ed on December 30, ty?quentlydetermined that pressure differences M Y' operation of one or more RCPs.yould have resulted in the P/T limits for both Units 1 and ? beihg exceeded in Mode 5. d Q AWA d- :2L kJ ADA1yab- of 0 cytt.gagit W Dackground (4 2 15; + u 'If POPS protects the RCS from exceeding the TS presR regenparaturc. (P/T) limits for plant heatup (reference TS Figure 3.4-2) and cooldown (reference TS rigure 3.4-3) by opening the Power Operated Relief Valves (PORV) during low temperature overpressure (LTOP) transients (RCs cold leg temperature below 312*F). Per existing design bases, either of the two PORVs has adequate relieving capacity to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle PCP with the secondary water temperature less than or equal to 50*F abovo the RCS cold leg temperature (heat addition), or (2) the start of an SI pump and resultant injection into a water' solid RCs (mass addition). The pressure limit at the low temperature end of the P/T curves is presently 450 psig for Unit I and 475 psig for Unit 2, as read fron the current heatup and cooldown curves. 3="5 Orf: 0=== w-514 was amorevad by the une en O 2I13/9E to th= f ar M10; ;;@. 1 -r-ed - :: th*--allows _.a-let r eyi: 50-adde M lisii.e . The original salem POPS analysis, based on the LOFTRAN computer code, calculated a maximum peak pressure for the most limiting mass addition transiant of 446 psig with the PORY set at a pressure of 375 psig. In this analysis, the RCS pressure due to injection,780 gpm sI flow into the initially cold water solid RCs Y
i i-07-1995 3rdd M FROM PSLC ODEP LICCWitC 609 339 5435 P. 6 j g. LICENSEE EVENT REPORT (LER) TEKT CONTINUATION Salen Generating Station Docket Number LER Number page Unit 1 50-272 94-017 4 of 5 hhalysis of Occurrencer leant' d) I was seasidered. In the limiting heat addition transient, a maalaun peak pressure of das psig was calculated with the PORV set at a pressure of 37s peig. The Nuclear steam supply system (NSSS) vendor identified in a_ , letter, dated March 15, 1993, a non-conservatism in the ; calculation for peak pressure for the heat input and mass addition transients that affects both Salen Units 1 and 2.- The - concern was that the difference between the wide range pressuro ' transmitters (PT403 and P7405) elevations, which sense het leg pressure and the reactor vessel midplane (where the Ts heatup aad ecoldown_p/T limits _are defined) Vith the RCPs operating van not considered in the original Salem POPS _ analysis.. To quantif;' the effects, specific pressure differences associated with RCP operation were determined for one, two, and four RCPs operating. The results of these calculations provided values of 2s, 37, and 71 psig with one, two, and four RCPs operating, - respectively. A correction pressure of 2 0 psig was then added i to account for transmitter elevation differences not previously accounted for in the original' calculations. When considering the ' pressure differential from one RCP in operation (31 psig) and adding this pressure o the paa pressure for the mass addition transient (446 peig the P/T its-for-both Salem Unita 1 and 2 are exceeded. Th ore, bo n units ware outside of-their design basis for reporting purp It was also determined that the P, if already in operation, would continue to operate upon initiation of a.SI signal if offsite power remained available. During this pcatulated event, letdown would automatically isolate as part of the SI actuation. The additional flow from the PDP is a concern for the limited period of tir.e when the RCS is </= 200*F (Mode 5), the PDP is in operation, and one (1) .CCP has its associated power supply available. The conbined flow-of 665 gpm from the PDP (105 gpr.) and-the CCP (560 gpr.) is now considered the most limiting mass addition transient. 5 PSE&G has re-analyzed this mass aadition event using the GOTHIC computer code assuming a bounding saximum pump flow rate of 675 gpm. The rasulting peak pressure is 474 psig, which exceeds the current-lizit of 450 poig for salem. Unit 1, but is within the current limits for Unit 2.
5-s*%1MS 3, G+1 FR31 PSEO OPER L!CCNS1tJG 609 33h Ed.15 P. 7 LICENSEE EVENT REVORT (LER) TEXT CONTINUATION 4 salem Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 5 of 5 Analysis of occurrence feon.t'dit For the heat addition transient (i.e., the start cf an idle RCP with the secondary water temperature </= 50*r above the RCS cold leg tenperature), the peak pressure is 449 psig, below the POPS limit of 450 psig for salem Unit 1 and 475 for salem Unit 2. Additional margin on the TS P/T curves can be obtained when operating with POPS (RCS cold legs </= 312'F) by NRC approval of ASME Code Case N-514. The code case allows exceeding the P/T limits calculated in accordance with 10CFR50, Appendix G, by 10%. ' As compensatory action for Unit 1, administrative controls ensure ' that Residual Heat kenoval (RHR) relief valve RH3 is available and the associated RHR isolation valves are in the open position. PSE&G has determined that RH3 has similar relieving capacity to that of a PORV. Tais action is only ne:ersary when RCS s temperature is </= 200'P, the PDP is in operation and a CCP has power available.
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Existing plant operating procedures for both Salen Units require that the power supplies to both 52 pumps be removed upon entry into Mode 4 (< 350 degrees F). Procedure revisions were implemented to limit the number of RCPs in operation to 1 while in Mode 5 (< 200 degrees F). On December 22, 1994, a 10CPR50.39 Safety Evaluation was completed for salem Unit 2 that changed the TOPS Technical specification Bases. The mass addition flow rate assumed for the present POPS analysis is limited to the combined flow from the CCP in conjunction with an operating PDP or SI pump while in Mode 5. On February.8, 199A, a 10cTR50.59 Safety Evaluation was completed for salem Unit I that changed the POPS Technical specification Bases.. The muss addition flow assumed for the POP 8 analysis was limited to the flow from a single CCP while in Mode 5. Following approval of the AsME code case N-514 by the NRC, an additional 50.59 safety Evaluation was completed that again changed the Technical Specification Bases. The new mass addition flow rate assumed for the POPS analysis is limited to the combined flow from the CCP in conjunction with an operating PDP or si pump while in Mode 5. _This is identical to-Unit 2 Bases. L* _ _ - - m__ __. . _ _ -
. . . _ . - . _ _ - - - - . - . . - . - - - - - . - . . _ . - _ _ = . . . - _ _ - _ - -
- P. 6 a-0* 1995 1 i 45Pt i FROi PSEO OPER L1CD4$1NG 609 339 5435 o
l LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ' Salas Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 6 of 5 Apearent cause of Occurreneet i This event is attributed to
- Design *, as classified in Appendix D 1 '
of UURIG-1022. This occurred because the NSSS vendor had not considered either the pressure differential associated with the I operation of the SCPs or PDP operation as part of design - basis analysis for the mass addition transient. j[c, P8E&O failed to recognize that both salen Units ve outside their design bases in December 1993, when considering the pressure differential associated with one or more Rcrs in operation. Prior similar Qccurrancet No-other prior similar occurrences have been identified related , to this design deficiency. ' EAfety sienificance1 This event is reportable in accordance with the requirements of 10CFn50. 73 (a) (2) (ii) (B) , due to the POPS not being able to nest
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3 its current design basis. This event had minimal safety - significance, based upon the additional relieving capaclty
.available through the use of RH3 and/or with the 10% allowance.
parmitted-by use of code case N-514. The new limits provided by the additional 10% margin allowed by Code Case N-514 ensures that the-current P/T limits for salon Units 1 and 2 will not be exceeded, based on either the original peak pressure calculated by the LOFTRAN computer code or the peak pressures calculated using the GOTHIC coDputer code. corr _ective Actions i The following administrative controls are in place on salem Units 1 and 2 to ensure compliance with the POP 8 analysis Lc
- 1. Oper ting procedures limit RCP operation in Mode 5 to pua / #
- 2. Power must lHe removed from the SI pumps upon entry into Mode 4 (350'F >Tave, > 2 00'F) .
_ ...m..~____._ _._._.._.._._.__ _ _ _ _._ ___.._._._. . . _ . _ _ _ _ . _ g-pi- K95 3 : 46pH FROtt PSEC. OPER LICENS1tc 609. 339 5415 F9 i 5
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L7CENstt EVENT REPORT (LEA) TEXT CONTINUATION sales. Generating Station to=ket trumber LER { Number Page ' 4 Unit 1 50-272 94-017 7 of 5 cerrantiv hetlen feont' din A submittal was made=to the NRC requesting permission to utilise AsNC Code Case N-Sid to allow an additional margin of 10% in-the P/T limits for the POPS during LTOP conditions. The NRC approved Pst&G use of code Case N-514 on February 13, 1995. / o ine corrective Action Program has been significantly improved with the issuance of Nuclear Administrative Procedure NC.NA-AP.25-0006(Q). Personnel involved have received reinforcement on procedure compliance responsibility for compliance with licensing commitments and probles reporting. Guidance has been provided to appropriate Engineering personnel regarding A n t Code application. , Management has re-emphasised the use of 100FR50.59 and its applicability to calculations / evolutions that affect design basis, basis of analysis, or conclusions in the UFSAR. Clay C.-Warren General Manager Sala:n Operations MJPJ g RErt-SORC Mtg. 95----
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%. n. emitn Principal Engineer - Nuclear Licensing FFOM: K. M. O'Gara Licens:ng Engineer
SUBJECT:
STATUS OF ATS ITEMS WHILE ON VACA" ION
.M .~r, . ., ,
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.i o. n. 1 While I will be on vacation between June 3 and 13, 1994, the following ATS tasks will be coming due that may require followue:
Generic Letter 92-01, Peactor Vessel Structural Integrity, ATS lter No. URC/GL/92-Ol/Tash 15 - Assigned to J. Perrin to provide Og'O{ ! talked :nput response by 6/06/94 for letter due to the Nhc on C/20/94. to J. Ferrin and he does net believe that he will make
\,theE/Eduedate, but should have draft input by E/10. This requ res tha: the letter be complete for SLB signoff on 06/15/94.
- is recommended that seteone be assigned to follow this task ::
go(d ensure that -he inte na'. ;ommitment to have rhis to SLB 3 worung d3ys ce:CIe 1:s 309 W1.. De mes. Far; ;'. 94-0032, H gh Fream Line Flow S! Instrumentation - Eza.uatacn cy NEE scheduled to be completea on nf/10/94 nl. Fre?:nese;. Le :er to NLF is in for signature w :n _. ha;kowski.
!%e /^ aay clerk enpires on June 29, 19?4 so should be able ::
ci:se th.s :ssue su: when : get back. Far: : N-M05, Moore Products Ecoster Relays - Evaluation by
. NEE s :neau.ea to be completed on 6 /1 ? / 04_ IV. Fregonese). Letter
- NLF .s :n fcr signature with L. ha nowski. The 60 day clock e::p.res en Ju.y ;, 1994 so : snould be able to close this issue
;ut .enen : ge rack.
The other Par 21's currently in process (94-0004, PORV Plug & Ste:n Material, and 94-0006, American Warming Backdraft Dampers) are due from UME by 06/20/94_and 07/01/94, respectively. Aga n, I should be able to follow these issues when : get back. The E day clock for 94-0004 capires on 7/09/94 and 94-000C enpires on
,/A-1i , a i o, 4. e The Letter to the NRC transmitting calculations in support of th+ ,Jy' degraded grid voltage LCR (93-10) has been s gned by J. Hagan.
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'The supplemental response to GL 92-01 for Hope Creek on Fluence ggy' Levels.is with you forfyout signature. This needs to go to SLB
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and should be given some priority. .
-j I have still been trying toimeet with Pete Ott to discuss the !
- Charging Pump _.,CR. He' cancelled-- ,e last meeting we had arrangea and was. working the back shift last week. This weeX he's been in ,
Pittsburgh. -The folder-is on my. desk. Maybe R. Villar could ! meet with him to finali:e theLchange. If not, I'll resolve it :
- when I return. ;
I'll also take care of the POP 3 setpoint isaue when I return. 1 However, I would still like to discuss the open issues with H. , Berrick before I leave (i.e., 450 vs. 450.7 man, pressura i
. following an inadvertent SI, and the need to perform an analysis l L
cf the Inadvertent SI,_with-a bubble in the pressuri:er and 1 kCP l-in service). If yLu'need to reach me, I'll be unavailable. - Only Kidding. See you in a couple of weeks. f L M e { 'F E D f k' i i ow i 3- }. I i u P F - re - p + t-- m e- %-- ------,--.,y y 9 - , , . . - - , , . . - . - , - # w, ei.3-~-,m, ,v 'rw *&- =g"=*ww -c' -- - 4* ""-' W s 4
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C NLR-N94193 United States Nuclear Regulatory Commiselon Document Control Deck Washington, DC 20555 Gentlemen: l l AdME CODE CASE N-514 SALEM GENERATING STATION UNIT NOS. 1 &2 FACILITY OPERATING LICENSE NOS. DPR-70 & DPR-75 1 DOCKET NOS. 50-272 & 50-311 Public Service Electric and Gas Company (PSF.&G) requests, in accordance with the requirements of 10CFR50.55a(a) (3) approval to utilize American Society Of Mechanical Engineers (ASME) Code Caso N-514. " Low Temperature Overpressure Protection", for Salem - Generating Station Unit Nos, i and 2. This letter also requests, in accordance with 10CFR50.11, " Specific Exemptions", an exemption from certain requirements of 10CFR50.60. 10CFR50.60 states that all nuclear power reactors shall meet the fracture toughness and material turveillance program requirements a for the reactor coolant pressure boundary as set forth in Appendix G and H of 10CFR50. Proposed alternatives to the requirements of 10CFR50.60 requires approval by the NRC in accordance with 10CFR50.12. The purpose of thia letter is to request approval of Code Case N-514 whien allows exceedance of the Pressure / Temperature (P/T) limits calculated in accordance with the requirements of 10CFR50, Appendix G by 10%. This additional 10% margin will increase the P/T limits operating margin for the Pressurizer Overpressure Protection Systcm (POPS) during Low Temperature Overpressure Protection (LTOP) conditions. This letter also addresses nonconservatisms recently identified in the POPS analysis. The initial POPS analysis for the mass addition transient did not account for (a) the differential pressure between the mid-plane of the core and the location of the pressure sensors located at the Reactor Coolant System hot legs with the reactor coolant pumps in operation, and (b) operation of the Positive Displacement Charging Pump following initiation of a Safety Injection signal. A cetailed discussion of the basis for the request, and technical justification for the exemption in accordance with 10CFR50.12 and 10CFR50.55a(a)3 is provided in Attachment 1.
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Document Cor. trol 7esh 2 NLR-N94193 Should you have any questions on this submittal, please contact us. Sincerely, J. J. Hagan Vice President - Nuclear Operations Attachment (1) Affidavit C Mr. T. T. Martin, Administrator - Region i U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. N. Olshan, Licensing Project-Manager - Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. C. S. Marschall (309) USNRC Senior Resident Inspector Mr. Kent Tosch, Managar, IV NJ Department of Environmental Protection
;1 vision of Environmental Quality Bureau of Nuclear Engineering CN 415-
-Trenton, NJ 08625
3 ._ Document Control Desk 3 NLR-N94193 KOG/ PEER REVIEW t BC Vice President - Nuclear Engineerina IN1. General Managar - QA/NSR (N46) Operations Manager - Salem (80) , Technical Manager --Salem (S02 Nuclear Mechanical Engineering
- car (N50)
Nuclear Engineering Sciences Mar.:A, '" 1 ) Manager - Nuclear Safety Review ' '.>) ' Manager - External Affairs (N2' Nuclear Electrical Engineering ura.ge t '
. ')
Onsite Safety Review Engineer 'a co 9 i Station Licenair,g Engineer - Sa no General Solicitor, R. Fryling, 0 r, SG) Mark J. Wetterhahn, Esq. B. O'Grady (Sol) G. Chen (S02) H. Berrick (N50) M. Dane,k (N50) V. Chandra (N32) Records Management, J. B. Caldwell (N21) Microfilm Copy File Nos. 1.2.1 (Salem), 5.38 ,
' Concurrences: For accuracy of information in your area af responsibility:
a Nuclear Mechanical Engineering Manager Date Nuclear Engineering' Sciences Manager Date i
. Manager Nuclear Engineering Design Date L ,
Doci. ament Control Desk 4
' NLR 419t 193 Concurrences: For accuracy of information in your area of s responsibility: j i
t i.
'lechnical Manager - Salem Operations Date ,
i Operations Manager - Salem Date
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REF: NLR-N94193 l STATE OF NEW JERSEY )
) SS.
COUNTY OF SALEM ) J. Hagan, being duly sworn according to law deposes and says: I am Vice President - Nuclear Operations of Public Service. _ Electric and Gas Company, and ao such, I find the matters set forth in the above refererced letter, concerning Salem Generating Station Unit Nos. 1 and 2, are true to the best of my knowledge, information and belief. i
-Subscribed and Sworn to before me this day of , 1994 Notary Public of New Jersey My Commission expires on 4
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NLR-N94193 ATTACHMENT 1 BACKGROUND The Technical Specification Pressure / Temperature (P/T) limits for plant heatup (Figure 3.4-2) and cooldown (Figure 3.4-3), which 1 are determined in accordance with the requirements of 10CFR50, Appendix G, ensure reactor vessel integrity. The 'eressurizer Overpressure Protection System (POPS) protects the Reactor Coolant System (RCS) from exceeding the Tech. Spe!. limits by opening the Power Operated Relief Vnives (PORV) during cold l overpressure transients (RCS cold leg temperature below 312*F). According-to the existing design bases, either PORV has adequate relieving capacity to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature (heat addition), or (2) the start of a safety Injection (SI) pump and resultant injection into a water solid RCS (mass addition). The pressure limits at the low temperature end of the P/T curves are 450 and 475 psig for Salem Units 1 and 2, respectively, as read from the current heatup and cooldown curves (Tech. Spec. Figures 3.4-2 and 3.4-3, respectively). The original Salem POPS analysis calculated a maximum peak pressure for the most limiting mass addition transient of 446 poig with the PORV set at a pressure of 375 psig. In this analysis, the RCS pressure due to injection of 780 gpm flow into an initially cold water solid RCS was considered. In the limiting heat addition transient, a maximum peak pressure of 418 psig was calculated. Westinghouse has identified in letter PSE-93-204 dated March 15, 1993 (NSAL-93-005B) a non-conservavism in the calculation for peak pressure for.the heat input and mass audition transients that affects both Salem Units 1 and 2. The concern is that the difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg pressure and the reactor vessel midplane (where the Tech. Spec heatup and cooldown pressure / temperature (P/T) limits are defined) with the reactor coc'snt pumps (RCP) operating was not considered in the original SaJem POPS analysis. This results in encroachment on-the P/T limits. To quantify the effects on Salem, specific pressure differences associated with RCP operation have been calculated for one, two and four RCPs operating. The results of these. calculations provided values of 29, 37, and 71 psig with'one, two and four RCPs operating, respectively. A correction pressure of 2.0 psig _ = 1
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NLR-N94193 ATTACHMENT 1 (Cont'd) has been added to these values to account for transmitter elevation differences not previously accounted for in the calculations. As a result, procedure revisions have been implemented to limit the number of RCPs in opera: ion to 1 while . in Mode 5 4 200*F). In addition, plant operating proceduren require that the power supplies be removed from both SI pumps (675 GPM) upon entry into Mode 4 (< 350*F). Therefore, only the mass input from a Centrifugal Charging Pump (CCP) (560 GPM) is considered. The net results, assuming operation of a CCP and the pressure difference from operation of one RCP, are peak pressures below the specified limits in the Unit 1 and 2 P/T curves. Since the completion of these evaluations, it has now been determined that the Positive Displacement Charging Pump (PDP), if already in operation, would continue to opecate upon initiation of a SI signal if offsite power remains available. During this postulated event, letdown would be autom Lically a as part of the SI actuation. The additional flow froe e PDP is concern for the period of time when the RCS .s < 200*F (Moce 5),
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the PDP is in operation, and one (1) CCP ha it's associatjei power supply available. Therefore, the combined ow of 6 p6 from the PDP (105 GPM) and the CCP (560 GPM) is now considered t'.e most limiting mass addition transient. 4 PSE&G has re-analyzed the mass addition event using the GOTHIC computer code assuming a bounding maximum combined pump flow rate of 675 gpm. The resulting peak pressure is 443 psig. After including the pressure differential due to the operation of one RCP and the 2 psig elevation correction, a peak pressure of 474 psig is established. This pressure exceeds the design basis limit of 450 psig for Salem Unit 1. The P/T limit of 475 psig for Unit 2 continues to be met with relatively no margin. These results are summarized in the following table: UNIT POPS CALC. PEAK AP W/ 1 RCP CORRECTED T/S SETPoINT PRESS. RUNNING PEAK RCS PRESS. (PSIG) (PSIG) (. ELEV. PRESSURE LIMITS - BASED oN CORRECTION (PSIG) HEATUP OR MASS INFUT (PaIG) 2 0
- F/ f.
CASE W/ 675 cooLDOctN GPM FLOW (PSIG) S1 375 443 31 474 450 S2 375 443 31 474 475 2
NLR-N94193 ATTACHMENT 1 (Cont'd) For the heat addition transient (i.e., the start of an idle RC* with the secondary water temperature less than or equal to 50 F above the RCS cold leg temperature), the original peak pressure calculated was 418 poig. When the pressure differential due to the operation of one RCP (29 psi) and the 2.0 psig to correct for elevation is added, the peak pressure is 449 psig. This peak pressure remains below the POPS limits of 450 and 475 psig for Salem Units 1 and 2, respectively. The NRC has been notified of tha potential for the Unit 1 P/T limits to be exceeded based on the above in accordance with 10CFR50.72 and 10CFR50.73. As compensatory action, administrative controls ensure RHR relief valve RH ilable and the associated RHR isolation valves are in 'u open pos ion. This action is necessary when RCS temperature i less-than 2 0 degrees F, the PDP is in operation and a CCP ha ower ava able. Analyses for a mass addition transient up to 780
- __aming either 2 PORVs, or 1 PORV 2nd RH3 are available, have determined that sufficient relieving capacity exists to ensure that the current P/T limits will not be exceeded. This is not required if a vent path iF established instead of POPS in accordance with Tech. Specs., or the PDP is not in cperation.
Salem Unit 1 operating procedures limit operation in Mode 5 to one RCP, require that power be removed from the SI pumps in Mode 4, and require relief valve RH3 to be available to ensure the current P/T limits are met when RCS temperature is < 200*F. Similar administrative controls tor Salem Unit 2 are in place although RH3 is not needed to meet its P/T limits. However, without credit for the relieving capacity of RH3, relativ71y no operating mhrgin between the current P/T limits calculated in accordance with 10CFR50, Appendix G and the peak pressure during Jyfjpgp low temperature overpressure transients exists for Salem Unit 2. Code Case N-514 allows exceedance of the P/T limits calculated in accordance with 10CFR50, Appendix G by 10%. The application of Code case N-514 provides sufficient margin such that the inadvertent SI actuation that results in a mass input from both a CC) and the PDP will not result in any P/T limits being exceeded. Code Case N-514 also allows the operation of up to two RCPs when RCS temperature is < 200*F, and removes the Salem Unit i requirement for RH3. Therefore, PSE&G requests application of the additional 10% pressure margin allowed by Code Case N-514 to address these operational restraints, and provide additional operating margin between the calculated peak pressure and the P/T limits. 3
y b NLR-N94193 [ ATTACHMENT 1 (Cont'd) DISCUSSION > Additional margin on the Tech. Spec. curves can be gained when- : operating with POPS (RCS cold legs less than- 312*F) by taking : credit for ASME_ Code Case N-514. This Code Case states that the j
- LTOP systems shall limit the maximum pressure in the vessel to ,
, 110% of tre pressure determined to saticfy_-Appendix G of ASME l 3_ Section XI, Article G-2215. Crediting the Code Case will allow . the maximum allowable pressure (Tech. Spec. P/T limits) for POPS
.to be increased to 495 psig and 522.5 psig for Salem Units 1 and-(It is noted that the contents of Code-Case N-2, respectively.
514 have been incorporated into Appendix G of Section XI of the [ ASME Code in the 1993 Addenda to Section XI.) j Current POPS P/T limits produtA operational constraints by ! limiting the range available to the operator to heat-up and
,"cooldown the plant.- For example, the' operating window through -;
which~the coerator can heat-up and cooldown the plant is determined 1oy the difference between the maximum allowable pressure determined from ASME Eection XI, Appendix 0 and, the minimum-_ allowable pressure for the reactor coolant pump seals (i.e. , an RCP operating limit caf 325 psig) . - In the future, the - < lowering of the POPS setpoint to account for the differential pressure between the core mid-plane and the RCS-hot legs may be 4 ~ necessary to ensure that the P/T limits contained in Technical Specification Figures 3.4-2 and 3.4-3 would still not be exceeded. Should the setpoint.be lowered due to the P/T limits, , the-pressure surges associated with the start of a RCP during ! reactor startup and during the filling and-venting process that . requires the " bumping" cf the'RCPs may result in the unnecessary . opening of a POPS relief valve. The current' POPS setpoint of 375 -l psig and the RCP operating limit of 325 psig have in the past , resulted in unnecessary challenges to the POPS._ Should the POPS ' setpoint_need to be lowered, it would be expected that additional c unnecessary challenges would occur. In addition, as-the reacter pressure' vessel becomes more irradiated, the POPS setpoint. requirements can impose significant operating burdens on the plants. As the vessel embrittlement , levels-are increased due to irradiation, the allowable pressure determined in accordance with ASME Section XI, Appendix G, at a given temperature:is lowered.- This. reduces the operating window for-heat-up and cooldown-during low temperature operation.
.Again,-reducing the operating window further increases the clikelihood of additional challenges to the POPS relief valves.
f' =i s 4 4 i 4
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. 1 NLR-N94193 ATTACRMENT 1 (Cont'd)
The ASME Working Group on Operating Plant Criteria (WGOPC) , developed code guidelines te define POPS limits that will avoid ! certain unnecessary operational restrictions, provide adequate i margins aqainst failures, and reduce the potential for unnecessary activation of pressure rel.aving devices used for POPS (PORVs). The philosophy used by the WGOPC for developing l these guidelines was to ensure that the POPS limits are still i below the P/r limits during normal plant operation, but allows the pressure that may occur with activation of the POPS to exceed the P/T limits by a maximum of 10% provided acceptable margins are maintained during these events. This philosophy was to protect the reactor vessel from low temperature overpressure transients and still maintain the P/T limits in the Technical Specificatione applicable for normal heatup and cooldown in accordance with Appendices G to 10CFR50 and Section XI of the ASME Code. The WGOPC appl.ied deterministic and probabilistic analysis techniques for several different flaw locations and heat-up and cooldown rates to establish the conditions delineated by the Code Case. For consideration, there are several conservatisms inherent in the development of the ASME Section XI, Appendix 3 P/T curve calculations (Reference 1) that include:
- 1) The safety factor of 2 on the principal menbrane (pressure) stresses.
25 A margin facto's applied to RT in accordance with the requirementsofRegulatoryGu$de1.99, Revision 2 (eg,, 2-sigma margins are applied in determining the adjusted reference temperature).
- 3) The disregarding of increased mechanical properties of the reactor vessel which accompany material embrittlement (eluvated yield strength and flow stress).
- 4) An assumed flaw in the wall of the reactor vessel that has a depth equal to 1/4 the thickness of the~ vessel wall and a length equal to 1-1/2 times the vessel wall thickness.
5)' The reference stress intensity curves defined in Section XI, Appendix G are usedsto bound the dynamic crack initiation and crack arrest toughness. 5 m
A 4 NLR-N94193 ATTACEMENT 3 '*ont'd) These conservatisma support the use of ASME Code Case N-514 to allow setting tne POPS setpoint such that the Appendix G curves are not exceeded by more than 10%. BASES FOR EXEMPTION REQUEST In accordance with the requirements of 10CFR 50.60, PSE&G believes the requested exemption meets the criteria in . l 10CFR50.12 (a) (2) and 10CFR50.55a(a) (3) with specis; exceptions = as follows: { 10CFR50.12 (a) (2) (li) l Application of tne regulation in the particular circumstances we.uld not serve the underlying purpose of the rule or is not necessary to achieve che underlying purpose of the rule. The Technical Specification bases for the current POPS setpoint is to ensure that the P/T curves calculated in accordance with Appendices G to 10CFR50 and Section XI of the ASME Code are not exceeded. ASME Code Case N-514 recognizes the conservatisms inherent in the methndology for calculating the P/T curves for heat-up and cooldown. The Code Case N-514 also recognizes that a POPS setpoint could be established which preserves the safety margins while allowing riant operation such-that unr.ecessary opening of the POPS will be prevented. Utilizing the Code Case N-514 criteria by increasing the maximum P/T limits by 10% satisfies the underlying purpose of the ASME Code and NRC regulat' ions while continuing to ensure an acceptable level of safety. 10CFR50.12 (a) (2) (iiil Compliance would result in unduc hardship or other costs that are rignificantly in excess of-those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated. Administrative restrictions on the number of RCPs that can be operated when RCS temperature is below 200*F are currently in place to account for the delta-P from the mid-plane of the core to the location of the pressure sensor for POPS actuation. In addition, current operating procedures require that power be removed from the SI pumps upon entry into Mode 4 to preclude the possibility of an inadvertent pump start. Relief valve RH3 is requi red to be available to ensure the current P/T limits are met i 6
D i NLR-N94193 ( ATTACHMENT 1 (Cont'd) L h when RCS temperature is < 200*F for Salem snit 1. Without credit for the relieving capacity of RH3, relatively no operating margin between the current P/T limits calculated in accordance with 10CFR50, tppendix G and the peak pressure during low temperature overpressure transients exists for - Sc'_em Unit 2. Code Case N-514 allows exceedance of the P/T limits calculated in accordance with 10CFR50, Appendix G by 10%. The application of Code Case N-514 provides sufficient margin such that the inadvertent SI actuation that results in a mass input from both a CCP and the PDP pump will not result in the P/T limits being exceeded. Code Cast " - 514 allows the operation of up to two RCPs when RCS temperature is < 200*F,_and remove the Salem Unit i requirement for RH3. As the reactor pressure vessel becomes further irradiated, the POPS setpoint requirements can impose significant operating burdens on the plants. As the vessel embrittlement levels are increased due to irradiation, the allowable pressure determined in accordance with ASME Section XI, Appendix G, at a given temperature is lowered. This reduces the operating window for heat-up and cooldown during low temperature operation. Again, reducing the operating window further increases the likelihood of challenges to the POPS relief valves. The guidelines contained in the Code Case for P/T limits provide acceptable margin againnt crack initil". ion and failure in reactor vessels. 10CFR50.12 (a; (2) (v) The_ exemption provides only temporary relief from the applicable reguleric" and PSE&G has made a good faith efforts to comply with the regs , tion.
;= PSE&G requests that the examption be granted until the time.that the NRC approves the Code Case for general use by the industry.
In addition, PSE&G is pursuing other efforts such as including Residual Heat Removal system relief valve RH3 in the POPS Technical Specification for low temperature overpressure transients to alleviate the long term potential impact of the omission of the pressure differential from the initial POPS analysj, and provide additional relieving capacity. PSE&G is currently in compliance with the requirements of 10CFR50.60, and-it is PSE&G's position that a good faith effort has been made to comply with the current regulation. 7
n e , I 'v s
-NLR-N94193 ATTACHMENT ' ( Cc .it ' o ,
10CFR50.55a f a) (3) LApprovalfof.the_use of Code" Case N-514 ensures an, acceptable level of quality and safety, and compliance with the specified ,
-requirements of 10CFR50.55a would result in hardship or_ unusual difficulty without a compensating increase-in the level of quality and safety.
As: discussed above, the Code-Case provides acceptable margin against crack initiation and failure in reactor vessels.. The Code case also reduces the potential for'the' unnecessary opening ', of the POPS.. Tnerefore, application!of the Code Case for Salem Units.1,and 2_ continues to ensure an-acceptable level of quality and' safety as discussed above in response to' 10CFR50.12 (a) (2) (iii) .
SUMMARY
ASME Code Case N-514' allows setting the' POPS setpoint such that
!the present-Technical Specification heat-up and.cooldown P/T
. curves in:accordance with Appendices G'of110CFR50 and Section'XI
-of the ASME' Code are not exceeded by more than 10% during low temperature. overpressure-transients. The ASME Code Committee has-concluded that Code 1 Case'N-514 provides acceptable margin against crack initiation andivessel failure. -The additional 10%-pressure
; margin allowed by Code Case N-514 removes existing operational restraints,_and provides additional-operating margin between the Lealculated peak-pressure and the P/T limits. Approval of-this
~ Code 1 Case also reduces-the potential for unnecessaryoPOPS actuations. Consequently,_ POPS limits ; allowed by Code: Case N-514 provides acceptable-margins'of safety while-providing additional.
operational-flexibility. i
- REFERENCES
- 1. -WCAP-14040, "Mcthodology Used to Develop Cold Overpressure
' Mitigating-System Setpoints and RCS Heatup and Cooldown Limit Curves", June, 1994.
8
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- NLR-N94193 .
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l37o-t;nited States Nuclear Regulatory Commission / Document Control Desk ,I Washington, DC 20555 Gentlemen: ASME CODE CASE N-514 SALEM GENERATING STATION UNIT NOS. 1 &2 FACILITY OPERATING LICFMSE NOS. DPR-70 & DPR-75
' DOCKET NOS. 50-272 & 50-321 Public Service Electric ~and Gas Company (PSE&G) requests, in accordance with the require ~ments of -100FR50.55a (a) (3) , approval to utilize American Society Of Mechanical Engineers (ASME) Code CaselN-514, __"
Low Temperature Overpressure Protection", for Salem Generating Station Unit Nos. 1 and 2. This. letter also requests,- in accordance with 10CFR50.12, " Specific Exemptions", an
' exemption from certain requirements of 10CFR50.60.
10CFR50.60 Jtates that all nuclear power reactors shall meet the-fracture toughness and_ material surveillance program requirements for the reactor coolant pressure boundary as set forth in Appendix _G and H of-10CFR50. Proposed alternatives to the requirementstof'10CFR50.60 requires approval by the NRC-in accordance with 10CFR50.12. The: purpose of this letter is to request approval of Code Case N-514 which_ allows exceedance.of ,
-the Pressure / Temperature (P/T) limits calculated in accordance with the requirements of 10CFR50,: Appendix G by 10%. This-additional 10% margin will. increase the P/T--limits operating margin for the Pressurizer Overpressure Protection System (POPS) during Low Temperature Overpressure Protection (LTOP) conditions.
This letter also addresses nonconservatisms-recan'.2.y identified by Westinghouse in the POPS analysis. The initial POPS analysis for the mass addition transient did not account for the differential pressure between the mid-plane of the core and the location of the pressure sensors located at the Reactor Coolant System' hot legs with the reactor coolant pumps in operation. A detailed discussion of the-basis for the request, and technical justification for the exemption in accordance with 10CFR50.12 and 10CFR50.55a(a)3 is provided in Attachment 1. u
P4
. Document Control Desk 2 NLR-N94193 Should you have any questions on this submittal, please contact us,
-Sincerely, J. J. Hagan Vice President -
Nuclear Operations Attachment (1) Affidavit C Mr. T. T. Martin, Administrator - Region 1 - U. S. Nuclear Regulator / Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. N. Olshan, Licensing Project Mane.ger - Salem U. S. Nuclear Regulatory Commission One-White Flint North 11555 Rockville Pike Rockville, MD 20852
'Mr. C. S. Marschall (SO9)
USNRC Senior Resident Inspector Mr. Kent Tosch, Manager, IV NJ Department of Environmental Protection-x Division of Environmental Quality Bureau of Nuclear Engineering , CN 415 Trenton, NJ 08625
Document Control Desk- 3 NLR N94193-KOG/- <
. PEER REVIEW .
BC Vice President - Nuclear Engineering (N19) General Manager - QA/NSR (N46) Operat. ions Manager - -Salem (Sol) Technical Manager . Salem (S02) Nuclear Mechanical' Engineering Manager (N50) Nuclear Engineering-Sciences Manager-(N32) Manager - Nuclear Safety Review (N38) Manager - External Affairs (N28) Nuclear Electrical Engineering Manager (N47) Onsite Safety Review Engineer - Salem (S12) Station Licensing Engineer - Salem (N21) General Solicitor, R. Fryling, Jr, -(Newark, SG) Mark J. Wetterhahn, Esq. H. Berrick (N50) M. Danak. (N50)- V.-Chandra (N32) Records Management, J. B. Caldwell (N21) Microfilm Copy File Nos. 1.1.1 (Salem), 5.38 Concurrences: For accuracy of information in your area-of responsibility:
~
Nuclear Mect.anical Engineering Manager Date Nuclear Engineering Sciences Manager Date Manager - Nuclear Engineering Design Date
Document Control Desk 4 NLR-N94193 Concurrences: For accuracy of information in your area of responsibility: Technical Manager - Salem Operations Date-Operations. Manager - Salem Date i I h
REF: NLR-N94193 STATE OF NEW JERSEY: )'
-) SS.
-COUNTY:OF_ SALEM )
J. Hagan, being=-duly sworn according to~ law:-deposes and says:- I1amLVi'ce President -LNuclear. Operations of:_Public: Service-Electric and Gas Company, and-as such,LI. find the matters set forth in the above referenced-letter,-concerning-Salem Generating Station Unit Nos; 1=and-;2,'are true to the best ot my_ knowledge, i.inf ormation- and-- belief .- Subscribed and Sworn to.before-me this: day of. ,- 1 9 9 4 -- Notary-'Public of-New Jersey-
-My' Commission' expires _on i.-
NLR-N94193 ATTACHMENT 1 BACKGROUND The Tech. Spec. P/T limits for plant heatup and cooldown, which re e ce miumu Acordance with the requirements of 10CFR50, ppendix G, ensure reactor vessel integrity. The Pressurizer Overpressure Protection System (POPS) protects the_ Reactor Coolant System (RCS) from exceeding the [rech. Soecl limits by pening the Powei Operated Relief Valves (PORV) during cold i f verpressure transients (RCS cold leg temperature below 312'F) . Either PORV has adequate relieving capacity to protect the RCS
\ A from overpressurization when the transient is limited to either
/ (1) the start of an idle RCP with the secondary water temperature ,
less than or equal to 50'F above the RCS cold leg temperature -g (heat addition), or (2) the start of a Safety Injection (SI) pump and resultant injection into a water solid RCS' mass a dition). W6f WM o
/q.The pressure limits at_the low temperature end 4are 450 and 475 68 , psig for Salem Units 1 and 2, respectively, as read from the
--} current heatup and cooldown 420T/h4 curves fech. Spec.\ Figures
; '3.4-2 and 3.4-3). The original Salem POPS analysis valudlated a E
[ maximum peak pressure for the most limiting mass addition transient of 446 psig with the PORV set at a pressure of 375
- r 't, psis In this analysis, the RCS pressure due to injection of 780 N { gpms flow into an initially cold water solid RCS vas y \s peak consioered. In the limiting heat addition transient, a maximum c3 pressure of 418 psig was calculated,with th: PCF" ::t 2t 2 Traccure ;f F ; i-L g pt.t.tn u thO g Westinghouse 4 1dentified in letter PSE 04 dated March 15, 1993
, (NSAL-93-005B) a =pcte.tial non-conservat 'sm in the calculation ef- (-
-e+re peak pressure f or - the ha.at input an mass addition transients
% that affects both Salem Units 1_ and 2.J The pr.a.s.su.r.a difference ,
I betweeg 10 C " ' ' ' g gwgog wh c1 range sensepressure transmitters'(PT403 hot leg pressure and the reactor and PT405) vessel d - w midplane (where the Tech. Spec. heatup and cooldown W pressure / temperature (P/T) limits are defined) with the reactor
'\b g Lcoolant pumos __(RCP) operating was not considered {1n theot p \
Westingnouse analysis ,, s___ _, , _ . y ,_ m , ,
~~Tn mantifv the_ Sal 7 ,, m_ ' ; .- -
p= v- _ - - specific pressure difference $ associated with RCP operatio na nasess the banan te of fe ~ - ti"~ SGs , thA Mitz-P de-calculated for one, two and four RCPs 3 operating. The re ults of the%alculation5 provide $delt P-y values of 29, 37, .ind 71 psig with one,J;wo and four_ RL y f F npe *m H nn . respe vive _ly. f ,n dfE__. w porrection maasure of 1
,/
2.0 psig wee,ad d to these dc1tr P-values to account for j transmitter ble tion differences 42ccucad in the calculations / Procedure vis ons have been im 'emented to limit the number of C@"\kj 1 0 poi- [lO \)
\g NM if .MwA hG trn
l hb" Ct6oN3; N {%N ftM %( LS be he 4bs.e sfe.m A m i,a 6 % NLR-N94193 . ATTACHMENT 1 (Cont'd) RCPsinoperatip*to1whileinMode5
.~....,,._..c._
(less than 200*F).
- M 2dmin4erratre: centrele 4n p1ac: sc limit,
-ly : RCP Jc cp..- a.. tL mois.ulc.E d pc2h peace"*e fer the /
/ m :: 2dditicr t. ono a ut i !?'.0 psig,.'n. : 599 ' --; the cairn 1=ema aoire n v31ue 5 ; ncp ;u yar;; im ,
_ __.__t- -
"= - - - _r l
^
's T 17 re it in e es e l' it at dhe Low e a e n f e ch cal Sp ifi tio hea p ad f
o w c ve fo Sa - .its ac 2 (<50 ig a nd 75 p si . r p i ) i e e- d n th ase ai d'ti on tra s'e ue o t, o ra ng
) o 1 eri g the pr_ an e _i ' er .
Tn t o.L w ng-Table s marizes c . resuics c Lthe abov dis : ss. n ba ed on the ri i nal POPS lysis f 'nject' of 780 pm SI f ow i oac 4 yter s id CS:
,, , n s \ /, , ,
3 / v (y s UN]w/ POPS [ CALC. PE , 4 / 1 CORRECTED
]T/SPRE SETPOINT PRESS. PEAK RC .,
(PSIG) (PSIG) R IN PRES iEATU R BASE ON (p - ) (PSI 2 0'F/){ / MAS 8 INDU'" OOOLDo N CASE W/> 80 GPM FL [' ' (PS$C)
/
S1 37h f[16l [ 31 I [.477 \ [M S2 M7!b 446 / 31 k/ 477 475 ff f.'e.tes k- / / lf r"-
*repgir eeri< eva:ua on and an _c hJ t h -
cons cvatis hat form .yse[s/ucrcrcquiredg the asesforgheo[O-ig' a yks y sana [ripP ysiqqco l d be reioved to d e t0:'ir if CH ~- rC2' i
- Oce nn = _ :pe rif c tien /T limits cou: d e met It was determi.ed tr at [h 2 SI mp flow ate c f 7E Og us d in t h oritgin POPS mpss add'-ion tran ient a- l i sis wa s xt c e l cor.ser ativea free ni al Sp cif' cation ur i lan Re< i ,ent
- 1. , . 2 identiiW s a m ximum f o rate for a pum and a Ce) t ifugal Charging Pump (CC ) as 675 gpm d 560 opm.
3 "e: une ectivelv g euu be cnt-plant operatingprocecurepregreythat
,* o, power supplies removed f rom both SI pumps'u,p'c6 entry into /
0 Mode 4; Thetetore, che inadverceuc wort of 2 C pump decc not /g k)(F [ nc; ts be cancidere d =e = -'er addit;cn tree.i e "rumption. og f rom 4$w? CCP 2nenm4 5r flcu re.t : of 60 gpm 45 g j) The mass input -" te b= C = g g a sd 63;a C o tSM - ! 4 P E g*WIdt%IESecIMIt7he Positive Displacement
.-_,, 1
, Charging Pump (PDP), if already in operation, would continue to operate upon initiation of a SI aignal if of fsite power ie- /th+1 available. Letdown would be automatically isolated as part of
- boJs 3% h {0,S k % h k ~Q N %
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.. k NLR-N94193 ATTACHMENT 1 (Cont'd) l Q g e. b
.whe efj actuation. !"he ^^P9 an21'/=4e p*^"4ded abo"e k=d ".e t e nciccred th; ddinvuol fiv- assuciated winn cae rvF alivuld a=
. Saf m Ly ec:i r ;ma:in. c- Lu une F0F5 to % red i to be cper: IcT The- additional flow from the PDP is enlap a concern for the period p1 t.inte when the o" is < 200*F g he PDP'is in - u d. :5~~
F operation,)and on_e (1) CCP has not had its assoc a CTd powerl __ g ;. F ~ -,.tne uvuEEned flow ot 665 GPM trom the upolv PHP removed.[d (105 uem an the CCP (560 GPM) is the most limi mass
'pa ^ -
addition transient,*ha* c-uld :: cur. MA g Thc ma;; additien d4Q ho e re-analyzed using the GOTHIC t r ien*a" computer code rea14er4 rally assuming a boun ing maximum co gined pump flow rate of 675 gpm. The resulting peak pressure m 443 psigy Whec.cco. Jc. ieg the pressure dif f erential due to 4.he-- (,' 7 vputation of one RCP, and the 2 psig elevation correctionj G4== 1
,_, e rh new Pnx m-eljsii? peak pressure of 474 psicf4xceeds h6.5 lM , the POPC limit of 450 psig for Salem Unit 1. The P/T limit of l
l 475 psig fo nit 2 we*M continues to be met with relatively no yNuA l- h margin. deresults of the nee POPC analysis baseu vu cliu mass l 1:; rat: cf 0 ; e.. - summarized in the following table: (f)l1.- ~ L-u UNIT POPS CALC. PEAK AP W/ 1 CORRECTED T/S SETPOINT PRESS. RCP PEAK RCS PRESS.- (PSIG) (PSIG) RUNNING PRESSURE LIMITE - BASED ON (PSIG) (PSIG) HEATUP OR-MASS INPUT e 20*F/HR k(\ CASE W/ 675 GPM FLOW gg g mA n,q 3 COOLDOWN (PSIG)
,3 t Y S1 375 443 31 474 450
'[ S2 375 443 -M 474 475 w.3, 'oteA
-The NRC wt.s notified of the potential for the Unit 1 P/T limits
-to be exceeded based on'the = -" rticn above in accordance with 10CFR503(E2R 94 -XTTh As compensatory action, administrative
- cvuu2cus ensure RHR relief valve RH3 is available and the associated RHR-isolation-valve is locked open with power removed.
g This action is necessary when RCS temperature is less than 200 d the PDP is in operation and a CCP has power availablo 0, h 9 [O3 egrees Analyses F,for a mass addition transient of T O GPM, assuming
'3 d either 2 PORVs, or 1 PORV and RH3 are available, have determined that t-hi- w hin=*4an W v=1vae per"ide sufficient relieving capacity.to ensure that the currey P/T limits will not be exceeded. This requirement "culd not ha,. required if a vent path t.L
-hm e been established instead of PODS in accordance with Tech.
Specs.. ( _ 3 _
e 4 ULR-N94193
. ATTACRMENT 1 (Cont'd)
For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature less than or equal to-50'F above the RCS cold leg temperature), the original peak pressure
. calculated was 418 psig. sure differential due to the operation of_one When RCP addin; (29,9hsitheand pres i2 $0 psig to correct for differ-n - i .- the tran;;itte- locaA) nn = = mmad in We elueho^ d celuulaL vs, the peak pressure is 449 psig. 9 r2ur; th: initial eendTtT L ---hi for the- heat- additic: tr2nrient are nc len;;r p* cent 2fter th; fitsi RCP is s cat uud, wuly the core delte-F due h tn _ n n a FP running ic =rplicable. This peak pressure h still O*-w.%
below the POPS limits of 450 and 475 peig for Salem Units 1 and d 2, respectively. I\ N a .- TI@~refo , IUI OP ana ysi ass ,in the omb d7a ac PD ) fYo' r e nd he f il e in i de he es e i e e C ot 1 g a d th m md l' a f e v sa 1 h ) r
@bo 1 ra ng t does no ; res t i l an .mm MFate oper 11:
y j Ok*** sue, salem unit I operatinig procedures limitene RCP in _y s rv e, require that-power be removed from the.s1 pump and /4 /lo require relief valve RH3 to-be available to ensure the cu ent P/T limits are met when RCS temperature is < 200'J Simila d( E0
- administrative controls for Salem Uni 2e@f4t'3_Ith$$ghRH3 1
%_ not needed to meet, M /T limits. & ,out t credit for the reileving capacity of RH3, relatively no operating margin betwee the current P/T limits calculated in accordance.with 10CFR50,
/4 Appendix overprescureG and the peak existp transients pressure during for low temp Salem-Unit 2. [erature F
Code Case N-514 ,/ff / 6_
+44a allowsexceedance of the P/T limits calculated in accordance with-10CFR50, Appendix G by 10%. The application of Code Case N-514 usea providq5 sufficient margin such that the inadvertent SI actuation that results in a mass input'from botn a CCP and the
'PDP_r""r will not result in dddi$b/T. limits being exceeded. Code
-Case N-514.will allowSthe operation of up to two RCPs when RCS
,- '~ temperature is < 200'F, and remove 5the Salem Unit 1 requirement.
xg34 for RH3. Therefore, PSE&G requests application of the additional 10% pressure margin allowed by Code Case N-514 to address these operational restraints, and provide additional operating margin between the calculated peak pressure and the P/T limits. DISCUSSION Additional margin on the Tech. Spec. curves can be gained when operating with POPS (RCS cold legs less than 312*F) by taking credit for ASME Code Case N-514. This Code Case states that the LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of ASME Section XI, Article G-2215. Crediting the Code Case will allow 4
NLR-N94193 ATTACHMENT 1 (Cont'd) the maximum allowable pressure (Tech. Spec. P/T limits) for POPS to be increased to 495 psig and 522.5 psig for Salem Units 1 and 2, respectively. (It is noted that the contents of Code Case N-514 have been incorporated into Appendix G of Section XI of the
'ASME Code in the 1993 Addenda to Section XI.)
Current POPS P/T limits can produce 3 operational constraints by' limiting the range available to the operator to heat-up and cooldown the plant. For example, the operating window through which the operator can heat-up and cooldown the plant is determined by the difference between the maximum allowable pressure determined from ASME Section XI, Appendix G and, the minimum allowable pressure for the reactor coolant pump seals (i.e., an RCP operating limit of 325 psig). In the future, the lowering of the POPS setpoint to account for the differential pressure between the core mid-plane and the RCS hot legs may be necessary to ensure that the P/T limits contained in Technical Specification Figures 3.4-2 and 3.4-3 would still not be exceeded. Should the setpoint be lowered due to the P/T limits, the pressure surges associated with the start of a RCP during reactor startup and during the filling and venting process that requires the " bumping" of the RCPs may result in the unnecessary opening of a POPS relief valve. The current POPS setpoint of 375 psig and the RCP operating limit of 325 psig have in the past resulted in unnecessary challenges to the POPS. Should the POPS setpoint neei to be lowered, it would be expected that additicnal unnecessar' challenges would occur. In addition, as the reactor pressure vessel becomes more irradiated, the POPS setpoint requirements can impose significant operating burdens on the plants. AG the vessel embrittlement levels are increased due to irradiation, the allowable pressure determined in accordance with ASME Section XI, Appendix G, at a given-temperature is lowered. This reduces the operating window for heat-up and cooldown during low temperature operation. Again, reducing the operating window further increases the likelihood of additional challenges to the POPS relief valves. The ASME Working Group on Operating Plant Criteria (WGOPC) developed code guidelines to define POPS limits that will avoid certain unnecessary operational restrictions, provide adequate margins against failures, and reduce the potential for unnecessary activation of pressure relieving devices used for POPS (PORVs). The philosophy used by the WGOPC for developing these guidelines was to ensure that the POPS limits are still below the P/T limits during normal plant operation, but allows the pressure that may occur with activation of the POPS to exceed the P/T limits by a maximum of 10% provided acceptable margins i 5
M NLR-N94193 ATTACRMENT 1 (Cont'd) are maintained during these events. This philosophy was to protect the reactor vessel from low temperature overpressure transients and still maintain the P/T limits in the Technical Specifications applicable for normal heatup and cooldown in ace:rdance with Appendices G to 10CFR$0 and Section XI of the ASME Code. The WGOPC applied deterministic and probabilistic analysis techniques for several different flaw locations and heat-up and cooldown rates taa escablish the conditions delineated by the Code 88 udt.rcM S A In addition, there are several conservatisms inherent in the development of the ASME Section XI, Appendix G P/T curve calculations (Reference 1) that include:
- 1) The safety factor of 2 on the principal membrane (pressure) stresses.
- 2) A margin factor applied to RT in accordance with the requirementsofRegulatoryGu!be1.99, Revision 2 (eg., 2-sigma margins are applied in determining the adjusted reference temperature).
- 3) The disregarding of increased mechanical properties of the reactor vessel which accompany material embrittlement (elevated yield strength and flow stress).
- 4) An assumed flaw in the wall of the reactor vessel that has a depth equal to 1/4 the thickness of the vessel wall and a length equal to 1-1/2 times the vessel wall thickness.
- 5) The reference stress intensity curves defined in Section XI, Appendix G are used to bound the dynamic crack initiation and crack arrest tcughness.
These conservatisms support the use of ASME Code Case N-514 to allow setting the POPS setpoint such that the Appendix G curves are not exceeded by more than 10%. BASES FOR EXEMPlION 2EQUEST In accordance with the requirements of 10CFR 50.60, PSE&G believes the requested exemption meets the criteria in 10CFR50.12 (a) (2) and 10CFR50.55a (a) (3) with special exceptions as follows: 6 I i
NLR-N94193 ATTACHMENT 1 (Cont'd) 10CFR50.12 (a) (2 ) (ii_)_ Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. The Technical Specification bases for the current POPS setpoint is to ensure that the P/T curves calculated in accordance with Appendices G to 10CFR50 and Section XI of the ASME Code are not exceeded. ASME Code Case N-514 recognizes the conservatisms inherent in the nethodology for calculating the P/T curves for heat-up and cooldown. The Code Case N-514 also recognizes that a POPS setpoint could be established which preserves the safety margins while allowing plant operation such that unnecessary opening of the POPS will be prevented. Utilizing the Code Case N-514 criteria by increasing the maximum P/T limits by 10% satisfies the underlying purpose of the ASME Code and NRC l regulations while continuing to ensure an acceptable level of safety. ( _10CFR50.12 ( a ) (2 ) (i ii) Compliance would result-in undue hardship ur other costs that are significantly in excess of those contemplated when the regulation was' adopted, or that are significantly in excess of those incurred by others similarly situated. Administrative restrictions on the number of RCPs that can be operated when RCS temperature is below 200'F are currently in place-to account for the delta-P from the mid-plane of the core to the location of the pressure sensor for POPS actuation. In addition, current operating procedures require that power be removed from the SI pumps upon entry into Mode 4 to preclude the _ possibility of an inadvertent pump start. Relief valve RH3 is required to be available to ensure the current P/T limits are met when RCS temperature is < 200*F for Salem Unit 1. Without credit for the relieving capacity of RH3, relatively no operating margin between the current P/T limits calculated in accordance with 10CFR50, Appendix G and the peak pressure during low temperature overpressure transients exists for Salem Unit 2. Code Case N-514 wedWL allowiexceedance of the P/T limits calculated in accordance with 10CFR50, Appendix G by 10%. The application of Code Case N-514.weth provide 3 sufficient margin such that the inadvertent SI actuation that results in a mass input from both a CCP and the PDP pump will not result in the P/T limits being exceeded. Code Case N-514 medi allowsthe operation of up to two RCPs when RCS temperature is < 200'F, and remove the Salem Unit i requirement for RH3. 7 i
r 9 I NLR-N94193 ATTACHMENT--1 (Cont'd) As the reactor pressure vessel becomes further irradiated, the POPS setpoint requirements can impose significant' operating 1
' burdens'on' ~ . plants. As.the vessel embrittlement levels are
> increased /due to irradiation, theJallowable pressure determined-in.accordancelwith ASME SectiondXI, Appendix G,-at afgiven:
__ temperature-isclowered. This reduces the-operating window for
-heat-up and cooldown during low: temperature operation. Again,-
-reducing the operating' window-further-increases.the-likelihood of
- challenges to the POPS relief valves. -The guidelines contained
-in the Code' Case.for.P/T limits provide acceptable margin against crack-initiation and. failure-in-reactor vessels.
10CFR50.12 (a) (2) (v) The' exemption wee 4diprovidqLonly temporary relief from the applicable regulation-and PSE&G has made a good faith efforts to comply with the regulation.
-PSE&G requests that-the exemption be granted-until-the time that-ithe NRC approves the Code Case for general uce by the industry.
In addition, PSE&G is pursuing other efforts such as including Residual Heat Removal system' relief valve RH3-in the' POPS-1 Technical Specification for low temperature' overpressure
-transients to alleviate the long term potential impact of tne omission of the: pressure differential from the11nitial POPS analysis and provide. additional relievingLeapacity.
PSELG is currently in compliance with the requirements of 10CFR50.60, and it_is PSE&G's position that a good faith effort has been made.to comply with the' current regulation. 10CFR50. 55a (a) (3 )' j ApprovalEof the use-of Code Case N-514'would'ensuresan acceptable-
' level of quality and safety, and compliance with the specified requirements of-10CFR50.55a would result in hardship or unusual
'difficulcy without a compensating 1 increase in the level of
- quality:and safety, i
'Asidiscussed-above, the. Code Case provides acceptable margin
- against crack initiation and failure in reactor vessels. The
. Code Case w&&&'also reduce 5the potential for the unnecessary opening-of thetPOPS.
- Therefore,; application of the Code Case for Salem Units 1 and 2 wouli continug te ensure an acceptable level of quality and safety as discussed above in response to 10CFR50.12 (a) (2) (iii) .
8 l
Ve -
- NLR-N94193 ATTACHMENT 1 (Cont'd) l SUMMMX ASME' Code Case N-514 allows setting the POPS setpoint such that-the present Technical Specification heat-up.and cooldown P/T-curves in- accordance with Appendices G of 10CFR50 and Section XI-r 'of-the-ASME. Code =are not exceeded by more than 10% during low temperature overpressure transients.- The ASME Code. Committee.has concluded that-code Case N-514 provides1 acceptable margin against crack 1 initiation and vessel failure, the additional.10% pressure margin allowed by Code Case N-514- woutd -removegexisting
- operational restraints, .and provide 5 additional operating margin between^the calculated peak pressure and'the P/T limits. Approval of this Code Case- eew+d also reducesthe potential for unnecessary POPS actuations. Consequently, POPS limits allowed by' Code Case N-514 provideJacceptable margins of safety while providing additional operational flexibility.
REFERENCES 1.WCAP-14040, " Methodology Ut -? to Develop-Cold Overpressure
- Mitigating System Setpoints at. o 'S ~ Heatup and Cooldown Limit l Curves", June, 1994 p
i 9 8
t , N f ( / [(M( p4m rea - TOs- T. K. Ross- / y/ M Nuclear Fuels Engineer - Salem S. Ketcham Principal Engineer -_ Nuclear Mechanical Engineering FROM: D.- A. Smith Principal Engineer - Nuclear Licensing
SUBJECT:
USE OF PRESSURIZER DORVS TO MITIGATE INADVERTENT SI AT POWER TRANSIPNT DATE: REF: NLR-I94553 On November 16, 1994, a me ing was held to discuss a draft-10CFR50.59 evaluation to rise the UFSAR. description of the inadvertent SI at power tr zient,-taking credit for the pressurizer-PORV's to prev '. subcooled liquid discharge-through the pressurizer safety val: s. Salem Technical, Nuclear Fuels, Mechanical Engineering, I&C Engineering and Licensing:were gt represented at the meeting. Mechanical Engineering-expressed y
-concerns relative to taking vedit for the PORV's on a permanent i basis (i.e., in support of a icensing' basis change), stating- tr #g N 9
-that PORV's were not historic lly credited for pressurc' relief,,. V' gi ,
and may not be fully qualifie for this type of. service.- A design change to trip the posit.ive displacement. charging pump on
'jjhvd.
an SI_ signal was discussed as a potential long term solution. y V ]..,-t This modification would increase the time to pressurizer filling N
-allowing more time for the operators to meet criteria in the Emergency Operating ProcedurestheSItermination/
k/j' . filling the pressurizer. (EOP's) prior to pg)V, p The purpose of this' memo is to document Licensino's evaluation of the suitability of taking credit for the pressurizer PORv's-to ,? l prevent subcooled liquid discharge through the pressurizer safety j-- .#- ' valves following an inadvertent SI at power. In summary. t[h*g , p , Licensing is aware of two issues affecting our-ability to credit J
' phe.Puxv's:
the capability of the air accumulators to support Dg PORV operation until normal control air is reestablished or the , N 't event is terminated,.and confirmation that the PORV qualification' @%. l g I testing sponsored.by EPRI includes tests which demongtrate the ability of.the PORVs to mitigate the inadvertent SI fransient, _ assuming continued operation of the positive displacement charging pump. Resolution of these issues should determine the appropriate course of action for the inadvertent SI at power transient (e.g., 50.59/UFSAR change to credit the PORV's, or an m
l i % o interim Engineering tvaluation). Details of Licensing's review is proc.ided below'. Safety Classification of PORVs and Associ3ted Ecuit, ment The MMIS component database was reviewed to determine the classification of the PORVs and support equipment. As noted in the attached' tables, the equipment necessary for PORV operation is classified as safety related, seismically qualified and subject to OA program requirements. Associated electrical equipment is environmentally qualified, with the exception of the PORV loss of air solenoid valves and the PORV loss of air pressure switches. Justification for exemp;ing these components from harsh environment EQ is contained in h.gineering Evaluation S-C-RC-CEE-0703. This justification is b sed in part on the fact that transients which require PORV opers :on, such ae Low Temperature Overpressure Protection (1,TO' I and Steam Generator Tube Pupture (SGTR), would not create a .ersh environment for the PORV' equipment. This justification woult also apply to the inadvertent SI event. NUREG-0737 Item II.D I. Performance Test no of Boilina Water Reactor and Pressurized Water Reactor Re),2f and Safety Valves The NRC's position as stated in NUREG-0737 ir " Pressurized-water reactor and boiling-water reactor licensees ano app'.icants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents." Salem participated Item II.D in an EPRI 1 required test program to satisfy this requirement. each plant to demonstrate the test conditions were equivalent to the conditions prescribed by the FSAR accident analyses. The NRC's Srfety Eva'_ cation Report (SER) for SGS conformance to NUREG-0737, item IT..D.1 wLs transmitted via NRC letter dated May 10, 1990 (reference 1). This SER documents NRC review of the EPRI/ Westinghouse test program. The program-included a review of the limiting transients relative to PORV operability for plants grouped together by type (e.g., Westinghouse four loop). These transients are described in EPRI NP-2296 (reference 2). Table 5-s#^ 3 of reference 2 documents the limiting valve inlet conditions f l associated with extended inadvertent SI at power events with Y liquid discharge. EPRI NP-2628-bR (reference 3) documents the tests performed on the PORV's, demonstrating ac7eptable performance of the Copes Vulcan 316 w/stellited plug and 17-4PH vcage design evaluated for SGS. NRC's' acceptance cf PORV and safety val've qualificatiop for the It gf SI evsht, however, did not rely on the test data. Inadvertentwas based on twenty minutes to fill the pressurizer allowing sufff:ient time to terminate the event before challenging the valves w2th subcooled liquid discharge (reference 1, par. 4.2.3). The twenty minute fill time wa. ased on failure of the PORVs to
- a
3:
~
. i l re i
open, and -was used to demonstrate the probability of liquid - challenges to the safety' valves was sufficiently low. SThis 20 minute ~ fill time is based on Westinghouse generic evaluations 3-
. (WCAP-10105) , and is discussed-in our letters to NRC dated August .
19, 1985 and March 31,1982 - (references 4 and- 5) . Reference 4 1 /*es states that' successful operation of.either PORV is sufficient te de's'g,I
-relieve the liquid supplied by both centrifugal _ charging pumps.
b
= Based on review of the EPRI reports (references 2 and 3), it in not clear whether the PORV: inlet conditions calculated for the.
inadvertent SI event accounted for operation of the Positive ' , .* " g Displacement (PD) Charging pump = flow. .The Westinghouse C-- 6 evaluation of continued:PE pump. operation, SECL-94-099 (reference.)V"# e( 6), evaluates the inadvertent.SI event assuming one PORV can provide sufficient water relief, but does'not. address any
) en Lpotential effects of the PD pump on PORV qualification testing.n s l
Nuclear Fuels is asked to determine whether the PORV conditions i _ associated with the current analyses are_ enveloped-by the EPRI
,3est-conditions. Mechanical Engineering could then determine '
whether the PORV test data supports valve qualification for mitigating an' inadvertent SI event. ~6D71. Tk4= anma_ question applies to PORV qualification' testing for the feedline break I accident and LTOP events, for which.NRC acceptance of NUREG-0737,i i II.D.1 was caseu on EPR1 test data (Reference 1). I, k Generic Letter 90-06 NRC's post-TMI review of PORV functions and qualification resulted in NRC Generic Letter (GL) 90-06. This letter identified three typical PORV functions. 1) For Steam Generator Tube Rupture (SGTR), the PORVs are manually opened to_ equalize the pressure-of the RCS and.the' faulted stehm generator. For SGS, pressure equalization is assumed.to occur within 30 minutes in the SGTR radiological dose calculcations (UFSAR 15.4.4.2) 2) Some plants, including SGS, use PORV's as the primary means of Low Temperature Overpressure Protection (LTOP). 3) PORVs may be used for decay heat removal to achieve plant cooldown ( e '. g . , following station blackout). GL 90-06 requested confirmation that the PORVs are subject to a QA program per 10CFR50, Appendix B, included in a maintenance and refurbishment program, subjected to Inservice Testing (IST), and have Technical Specifications with Limiting Conditions of Operation for inoperable PORV(s), i-PSE&G's initial response to the GL (NLR-NE0234, 12/21/90) committed to meeting the GL requirementr, including IST for the PORV air accumulator check valves, and submittal of g License Change _ Request to make the PORV and L'/OP Technical Specifications more restrictive. , By letter dated February 8, 1994 and SGS Unit 1/2 License Amendments 150/130, NRC accepted out position on GL 90-06. SGS
s i
'1cchnical Specificat:lons limii the 'allwable time <of plant operation with one or both'PORVS incperab'le and 2 cat capable of being manually cycled. As discussed belov, the capability to manually cycle the PORVs relies ot. the 4rcumusmrc in the event of loss of normal contrct1 win PORV . sir Accumulators The SGS PORVs require cont'rol air to oper. Air accumulators are provided=to maintair.the ability te cycle the PORVs in the~ event of it is of normal control air (e.g., following a Phase A y
h containment isol tion due to SGTR or LTOP cvent). Engineering f F.)% Evaluation S-C-CA-MDC-1169 eva]uates the pressure decay of the M accumulators, including an allowance for check valve leakage.
,eF gu The sizing basis for the accumulators used in the evaluation is e # '#5 l suf ficient air to fully cycle each PORV 100 times under LTOP conditions.
eg(, A 50.59-to credit the PORVs to mitigate the inadvertent SI event would have to address the adequacy of the PORV air supply, and should account for accumulator check valve leakage allowed by Inservice Testing. As pointed out by Mechanical Engineering
./ during our November 16, 1994 meeting, the air requirements associated with sustained PORV liquid discharges are difficult to
- determine because the PORV(s) would tend to chatter rather than fully cycle. However, this appears to be true of our present accumulator sizing basis (LTOP), although the LTOP fluid temperature and pressure conditions are much lower than those of g the inadvertent SI event.
Measures that may be taken to enhance the availability of control
,ir to the PORVs include procedural steps to take manual control of the PORVs and/or reopen the control air containment isolation valves (CA330's) and verify control air header pressure following reset of the SI signal.
References'
- 1) Letter from J. C. Stone (NRC) to S. E. Miltenberger (PSE&G),
" Safety Evaluation Report, NUREG-0737, Item II.D.1, p
Performat.ce Testing of Relief and Safety Valves," May 10, s 1990.
- 2) ' Valve Inlet Fluid Conditions for Pressurizer Safety and t Relief Valves in Westinghouse-Designed Plants," EPRI NP-2296, January 1983.
- 3) "EPRI PWR Safety and Relief Valve Test Report," EPRI NP-2628-SR, Decemoer, 1982.
- 4) Letter from C. A. McNeill (PSE&G) to S. A. Varga (NRC),
" Request for Additional Information, NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves,"
August 19, 1985. l
l m
)ne ,
5): ; Letter from-E.=A.-Liden:(PSE&G) to S. A. Varga (NRC), NUREG-0737, Item II.D.1,-3 PWR Safety and-Relief Valve Test Program,"~ March 31, 1982".-
- 6) ~ Letter'from J. Huckabee (Westinghouse) to E._S.-Rosenfeld l(PSE&G),.!PSE-94-657,-Safety Evaluation-for Revised ECCS
'FJows_due;to-PDl Pump _ Operation (SECL 94 -0 99) , August. -10, 1994.
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l Per MMIS, the following equipment is qualified as Class 1E*: 2PT455 Pressure Transmitter 2PO455 Loop Power Supply 2PM455G Signal Summator 2PM455G/R Input Module 2PM455A Signal Isolator ** 2PM455A/R Input Module 2P455A Signal Conditioner 2HC455C Manual / Auto Station 2PC455K Controller - demand signal Iar 2PR2 2PC455K/R Input Modules 2PC455K/R1 2PC455K/R2 2PI455D Pzr Press. Control Channel I indicator berel 2PC455E Comparator - Hi Pressure /1PR2 interlock 2PC455E/R Input Module i
- Selected components are ass;;iated with control of 2PR2; analogous componenets for 1PR1, 1PR2 and 2PR1 have similar qualifica* ion.
** Isolates PORV control circuitry from reactor protection system.
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C99tISBNT NCBIBER Manamy sea w._;,_smens asiega ir ~ SECTICat I (Initistad .--pcPs sets.1st = 5 ee m tise 23P03T susync"rs .24.fgi /.3_. TI3EE! 5/ 66 I
'ORIT (81,83,83,EC) s1 a 12 DATE OF INCIDENTt St20EnRY OF EVENT (IF ESF ACTUATION, INCLUDR SOE PRINTotFT) t i f he POPS
.n a lottar on 3/15/93. Westmohouse identifisc a rencenservatism in their heat which effected both utzts, The terecenservatism was evaluated by NME.
93 917 (12/30/93). Sub.tequently, it w as in MEC-Case N 614 used m MEC-43-917. d 2 peig (Um determined that unti approved by the NRC. Salem could ret teks credit fo 2 with a 478 pois limit). These consstions were not reported to the NMC at snat time. Additional caiculations were made by Wettachoose MEC 94 630 Unit't could exceed its DEF A94-0080 ws written to track the overpressure issue 15126/961 reported the res ' prasst a lamm 8450 psag) by 0 7 psig during a mass sedition traris ant Delewt 200 safetyf. T On 11/17194, an IR was issued to document a further i MEC discove*y 94 630. ;nduated thatinatIM UrutPDP2 wasmay be runmng d 1raection at now temperature, in house calcu!ations t'y 8 NES simdar to f Unit t to used t'ioss adoress the PDP issue nposttien I i when RCS acceptabte, but Unit ) exceedad its pressure limit by 24 pts. LER 272/ 4
- temperature is s200'Fl. T emits for rae i
Ccds Case N-514 was approved for PSE&G use an 2/13/95. This f LTOP provided conditions. ads.t enet ma POSS et. ring LTOP conditions! Altnough inis now puts Unit 1 and 2 orerst.ons 5 25/94. Reference witnr. tN ces4an taa Attachments PSE&G stil reeos to report operations outside the design and neensir.g tsses on 12t30/93 enc / - 14. 1 340 - usa /0$R PHONE EXT: - t DEPT: - L c . tut or _ 4/" REPORTED BY i
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% UNIT LOAD: e,; d DSte EX PWR AT TDIE OF EVENT:
1 (IF YES, ATTACH ECO COPY) REPORT MhDE PER ECG7 (Y/N): _ W.R.#: DATE IN: _ TIME IN: _ LCO # , _ -A/S # f ERGolatEL PROCEDURAL ,,,,,, INITIAL CAUSE DETERMINATI0Wt EQUIP, ,_,. DESIGN OTHER: @"' 'ub -
-REPORTABLE: h /NO, * *0F /$K 97 C % NM DAT33 - / /I c ^_
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' pressure pressure +194 Peig In POPS traneeltter palg E 'e m
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Analyele ' 2 4 MCPS 2 BCPs 4 Polg Mcps RCPs (pel) (pel) Pois pelg 4,5 72 39 51s ess 45e
., Unit 1 375 4%
ass 475 522.5 72 3s 519 375 446 [;I unit-2 .
. I, .
- I a
roe c-93-9n , haw 12 /solr 3 l y: . y'gwcs : pops up. + n.~~~s ~
- s-o Berrscxi,sa,,ar :
y
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l
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, TABLE 1
;t*
DELTA P (VESSEL STATIC HEAD CORRECTED LTOP ALLOWAst,E
'_ ,' .LEM ESTIMATED. INITIATING EVENT MID PLANE TO TROM PEAK PRESSURE PRESSURE EDR
;g IIT. PEAR PRESSURE FOR LTOP TRANSMITTER IF TRANSMITTER TO AtTER LTOP CONDITION
'" StBER IN PSIG FOR VESSEL MID ADDRESSING TO MEET ArrEMOIx~
MASS (M) & IBASED ON TECH APPLICABLE Q MEAT (H) . SPEC BASIS { BASED IN CORE PLANE IF wESTINGMOUSE G CRITERIA
*" 3/4.4.9 FOR UNIT DELTA P APPLICABLE NSAC-93-05B IN INPUT CASES PSIG [ EASED ON TECM '" 1 s 3/4.4.10 ft)R CALCULATED BT (TROM SPEC NEAT UP AND sGS/M/DM 42 UNIT 21 WESTIF-HOUSC: & (NOT TO BE i "' BASED ON ,%DDED NMERE (SLD( OF COLUMN COOL DOWN CURVES AND 62 JELTA P FROM 92 AND COLUret i
PSOWIDED IN T.S. III PESPECTIVELYj OPERATION OF ONE RCP AS PREVIOUS COLUMN W4 OR W5 AS FIGURES 3.4-3 PROCEDURALLY IS ADDED APPLICABLE) AND 3.4-4I .
~ * '
- RESTRICTED IN ALREADYI
*** MODE 5) f 7 .
N/A 9 7 'M IN*JS I 459.7 PS 450 PSIG
-* 446 PSIG THE START Or A C '~' NIT I SI PUMP AND ITS 8 6 ' -2" == 10. 8
- MASS INPUT INJECTION INTO A - 4.7 PSIG L_
h ~ WATER SOLID RCS
, (No RCP RUNNING) 449 PSIG 4',G PSIG START OF AN IDLE 29+2= 31 PSIG N/A NET 1 418 (29 IS ET)R I .- HE)*T INPL'T PCP NITH THE 1 SECONDARY WATER RCP FROM PSE-TEMP OF S/G LE' $91-707 AND 2 f THAN OR EQUAL 20i PSI WAS ADDED * *: 1 "* 50 F ABOVE PCS j TO CORRECT ,.,, ' OLD LEG TEMP TRASHITTER El.
OF 92.4' TO HOT [ I.EG EL O F 97 ' I .
'i
. . i. 4 a ~
~
N u. un,+ 1 .7 jossy e>wecclacee, . mEC -W -(,3o , dhd o S/z.4/79 1 m 5 fe 1 c. ace : k-h k rict 6 w.cdernua 1
;C m a.c.nuc a,<,,, in foes Ju ro,af D
;( . ?4. ; &ip4 ;4:02 % T EC.- ( F 0,01 N
' O I V('.$ &% P49 U p .Ep e x /TM l4 i
) .. . . . . . ... .. . .... .. . . ------ -- RastARR8 / 00eouwr8 ------- - ------ - ----- -----TcNo s 2 4 l 10tAND'INFVT ===$ ,, PAGE 4 0F 4 DISPLAY N004 j PR NUMBERI __,940937136 EVAL TYPE: DF85 BVAL NBRt 01 l THE EVALUATION OF TME DISCREPANCY 18 DIVIDED INTO TWO SECTIONS FOR MODE 4 AND j NODE 9. IT Z8 CONCLUDD TNAT THERE IS NO OPERABILITY CONDERN CAUSED BY THIS _, j DISCRDANCY AND IT Don NOT INVOLVE ANY CONDITIONS THAT MAY 88 REPORTABLE._ l A. 11003 4 EVALUATION I j POLLONIMS DOCUMENTS NERE REVIEWED POR TilIS EVALUATICII: TECN. SPEC. ! 3.1.3.4, 3.4.18.3, AND 3.8.3, UFSAR 7.8.3, AND PROCEDURE 107-8. IN NODE 4, j IT Z8 REQUIRD To NAVE AT LEAST ONE CENTRIFUSAL LNARSING PUMP IN OPRABLE l CONDITION PER TECN. SPEC. 3.1.3.4 AND 3.8.3. SW TEE FOOTWOTE OF 3.8.3 ! RBOUIRES TNAT ALL SI PtStPS DE TASSED OUT IF RCS COLD LES Is ,313 DBS.F. _ _ _ i THIS PRSCLUDE ANY POSSIBILITY OF ANY INADVERTENT Ill3ECTION OF SI PUNPS WHICN. j .NAY CAUSE THE CONCERN ADDRESSED IN TNIS DEF. 48 $NONN IN 15E ATTACIOtIWT OF i THIS DEF, THE WOR 87 CASE FEAR FRE85URE CARSED BY AN INADVERTENT INJECTION OF_ j A CENTRIFUGAL CNARGING PUltP Z8 SEIANt THE TECN. SPEC. LIMITS. Til08 THE i DISCRDANCY IDENTIFIED IS NOT OF OPERATIONAL SIGNIFICANCE IN MODE 4. ! 3. MODE 8 EVALUATION ! NORE TECN. SPEC. SECTIONS ( 3 .1. 3 . 3 AND 3 . 4 .1. 4 ) AND PROC EDURES ( OP-PI . SJ j -003, OP-so.8N-003 di 003) NME REVIEWED FOR IIODE 8 EVALUATION. DURING j IMPLEMENTATON OF OP-80.5N-003 OR 003 POR SERVICE NATER READER OWAGE, WERE_,_ l DISP RVNat ! RVNis COMPS ! $8E88 AGE: REVIEW CONNENTS AND PF4 TO RETURN TO DISPOSITICH SCREEN.
- lY JOB LU #3 :
I i j . .... . .. . .. .. .. -------------- mENARx 8 / C0sonorTs -------------------------TcMo 3 4 i CONNAND INPW ===C PA02 4 0F 4 DISPLhY NODE PR NUMBER: _ 940937128 EVAL TYPE: DF8E JVAL NBR 01 i I IS A P088%BILITY THAT ONE SI PUMP CAN DE P! ACED IN OPERABLE CONDITION. THERR_ ! IS NO' RESTRICTION IN THE PROCEDURE TO PREVENT THE- OPERATOR FROM MAKIMO THAT_ l CNOICE. TNUS A POTENTIAL Or INADvERDENT SI INJECTION EXITS. N0amALLY TNE SI_
- j. PUMPS ARE TAFC AD OW AND Atfro INJECTION SIONALS ARE BLOCRED IN MODE 8. A PROCEDURE REVISION REQUEST NILL BE ISSUED TO CHANGE OP-80.sN-003 AND 003 ,
TO PRECLUDE THE Poss!81LITY OF SI INJECTION IF ANY RCP IS RUNNING SY , i RESTRICT!0N or CHOICE OF ANY WO SAFETY GRADE PUNP98 IN THE PREPARATION OF j SDVICE WATER NEADER' OUTAGES. THE REVISION WILL PREVENT ANY SI PUMP FROM __ ! sEING PLACED IN OPERAsLE CONDITICII TO MEET THE REQUIREMENT OF TSCN. 8PBc. ! 3.4.1.4. ! TEE CollCERN IDENTIFIED SY TRE 'WOUIA SE POSSIBLE' SCEllARIO 0F TRIS DEF_ j IS TNh2EFUL DET MNINED TO BE NO OPERATIONAL SIGNIFICANCE AND DOES NOT INVILVE
- ANY CONDITION THAT MAY BE RDORTABLE.
i BY G. C'!EN E 3784 l , i. DISP RVN2 a
- RVN1: CCILPs n' MESSAGE: No NORE DATA TO SCROLL IN PORNARD DIRECTION p' '
--..--,--,+.-..r,wem., .+---....-..----,w- w w rw v * *
.c-a4 usa .7:c3 ces335aus w.& n :- :E:' c.02 'l ATTACIIMENT11) PRODLEM REPORT #940927114 A De6clency Report [DEF. DES.90 0060) was imidated on April 19,1994 to address non i conservatiam in POPS set point to addeens the Westinghouse Nucleet Safety Adviscwy Letter PSE.93 204. The Westinghouse notincation identi8ed an error in their analysis when della P u a resuh of RCPs running at the elevadonal diferer.ae of the wide rasse pressure tronomitter to the mid plans of the r~ was not addressed. The 1: sue was '
resobred throush a letter kom H. terrick to Wiedemans, MEC 94430 dated May 26, 1994 The DFJ was subsequently closed usinh he results presented le the MBC.94 610. The issue wu evaluated to be not a concern based on the current plant LTOP dange baals u desenbod in the Teek Spes Duls 3/4.4.9 [ Unit 1) and 3/4 A.10 (Unit 2]. The current l plant ult)F deman basis is "an inadvertent start of a $1 pump ir(ecting into a water solid RC5 " The operadon of a RCP when RC$ is water solid, is precluded procedurally [$ OP.50.RC 000l(Q) ]. Because of this, the pressure d@srence associated with RCP operation was not added to previously calculated peak pressure ofinass input case, however, the hydrostatic had efect was added While pursuing to take credit for RH3 [RNR suction reliefvalve) for the LTOP analysis fu future Tech Spec emendment , a new worst case scenario wu identi6ed that could afect the LTOP est point in a non conservative way for a mus input case scenario The scenano is this: RCS is in mode 5 with a PZR bubble and a RCP running; inadvenant Si starts that eventually fills the PZR, with the RCP and SI pump still runningc Thle l , situation would resuh in the addition of the delta P unociated with the running RCP to the RCS peak p ensure In this situation, the corTacted peak RCS prissure would onceed the Tech Spec allowable pressure, as determined Som the limitmg cooldown curve The - results are tabulated below. ) POPS CALCULATED DELTA F TO CORRECT 1. TECit SPEC thrlT SET PEAN 'BEADDED D PEAX ACS FAsa$11AE POfMT PRESSURE IN FOR ONit PRESSURE LIMITS PSIG F58G RCP BASED ON [FOR TEE RETNNING i HEAT tJP OR LIMITING INMODES to dearse :'
- MASS INPUT 640F 6 ! F/BR CASE RA8ED ON LIMM l C00LDowN 760 GPM 81 MODE S i RATE
! FthWJ OPERATION j TO 1 RCPI L $1 375 446 31 477 '450 c S1 375 446 31 477 i 473 ItM 1 INPUT C4St 15 NOT Ukffn IN THIS TA RI.R AN li IN WOT !)M77/W(1 l L Although this new scenano is not a part of the current plant design basis, a conservativs interptstation was made to apply this "would be possible" scenario while evaluating the L Westin6 house concern This results in Salem _ calculated and corteesed peak pressere that I encoeds the allowable limus of hestup and cooldown curve. This problem report is generated to utentify this concern Ibr ibnhor action na may be required . 3 r
,_ -- ._ . . ~ , . . , , - , _ _ _ , . . , . , _ _ , _ . _ . , , , . + . . . , _ _ . . . . , , - . ~ . . , , ,
..,.-_....._.,_,-3.. .,-___-,.e-., , . , . . . . , , . , _ , . - . _ , - , . . , , -
e ic-x-tas4 pics en an 1'4e u e- Tsc- tre- 2.0 i SAFETY RIGNIFICANCE The current plant design basis uses 446 pais na the resultant puk RCS pressure which is based on 760 OPM as the $1 Aow. Since two $1 pumps and one Charging pump are tagged out when RCS cold les tempervture falls below 312 7 [10P4), the inadvertent Bow in this Mode la limited to $60 ppm, the runout flow of a Charging purnp. This more realistic maximum $1 Dow results in a calculated peak prepure [ sing OOTHIC Version 4 Code) as tabulated below and the reouks in the Table constude that the resuhant corrected RCS pressure remains bounded by the plant heat up and cooldown curves. trNf7 POPS CALCUtaTsp DELTA P CORRECTED TECESPEC TECESPEC - SET MER PEAX TO 38 PEAIC RCS PRESSURE 1ADT + POINT PRee8URE1N ADDED PRESSURE L01011 le %
- l P$lG PSIC POR ONE BASED ON ~PRENORE '
GDR TME ACF MEAT UP OR EsFBIG LDUTING REMfftNG MP/3RR PEphtitTED MARI INPUT rM WlODE S COOLDOWN SY ASME . CASE BASED RATI COG CASS ON He GPM $1 - NSH FthWJ S1 375 407 il 438 450 4 95
~ 52 375 407 31 438 475 522 5 Additionally, a recent calculation takmg credit for RH 3 suction relief to piovide LTOP mitigation has calculated the peak pressure for mass input case u 411 pois such that the corrected peak pressure aRer accounting for a RCP run delta P will not infringe on the plant beat up or cooldown curves As idemi6ed above and as further discussed below, the current infrinscment of plant hatup and cooldown curves for both Salem Unita do not prwent a safety signi6cance or
' impact to contmued operation of both Salem Unitt, PSEAG will pursue revising the current plant design basis addreasing the reduced $1 flow and also subsequently taking credit for RID relief valve through a Licenac Change Request. Howeser, in the immediate fvture, we will be requesting approval of ASNm Code Case N 514 that will provide relief by allowing up to 110 % allowable vessel maximum pressure to allenate the concern arising from the Westinghope Advisory Letter. The approval of the Code Case N$14 wiil alleviate the infringement of the heat up or cooldown curves with the current plant design basis without any reanalysis required-The current 10P 6 procedure limits the plant operation below 312 F to one charging pump and to one RCP when the RCS temperature is below 200 F This procedural aspect must not be chanted as it will have impact to the Salem plant's LTOP desian basis -
2 n -
, .% O s.
DESCRIPTION OF OCCURRENCE On November 17, 1994, Salem Unit I was in Mode 1 at 100% power. On that day, PSE&G determined that the pressure limit of 450 psig may be exceeded for Salem Unit 1 when considering the additional flow from the Positive Displacement Pump (PDP). The Technical Specification Pressure / Temperature limits for plant heatup (Figure 3.4-2) and cooldown (Figure 3.4-3), which are determined in accordance with 10CFR50, Appendix G, ensure reactor vessel integrity. The current bases for the Pressurizer Overpressure Protection System (Technical Specification 3/4.4.9.3) states that one POPS rel**f valve provides adequate relieving capacity in the event of ! rstepressure transient that includes the inadvertent start of a s pu,o (Mass Addition transient) into a water solid Reactor Cooitnt 1/ stem (RCS). PSE&G has determined that the following realistic mass addition transient assumptions when considering the additional flow from the PDP would place Salem Unit 1 outside the limiting design and licensing basis POPS analysis: RCS temperature less than 200'F
- One Reactor Coolant Pump (RCP) in operation
- PDP in service a maximum of one Centrifugal Charging Pump (CCP) has its power supply available Initiation of a SI signal would result in a mass addition from a combination of flow from the PDP and the CCP based on the above assumptions, and could result in a peak RCS pressure of 474 psig that exceeds the design basis pressure limit of 450 psig for Salem Unit 1.
ANALYSIS OF OCCURRENCE The Pressurizer Overpressure Protection System (POPS) protects the Reactor Coolant System (RCS) from exceeding the Tech. Spec. Pressure / Temperature (P/T) limits for plant heatup (Figure 3.4-2) anc cooldown (Figure 3.4-3) by opening the Power Operated Relief Valves (PORV) during cold overpressure transients (RCS cold leg temperature below 312*F) . The limits are determined in accordance with the requirements of 10CFR50, Appendix G. Acco ding to the existing design bases, either PORV has adequate relieving capacity to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature (heat addition), or (2) the start of a safety Injection (SI) pump and resultant injection into a water solid RCS (mass addition). 4 l
V The pressure limits at the low temperature end of the P/T curves ! are 450 and 475 psig for Salem Units 1 and 2, respectively, as ; read from the current heatup and coolduwn curves (Tech. Spec. 1 Figures 3.4-2 and 3.4-3, respectively). The original Salem POPS analysis calculated a maximum peak pressure for the most limiting mass addition transient of 446 psig with the PORV set at a pressure of 375 psig. In this analysis, the RCS pressure due to injection of 780 gpm flow into an initially cold water solid RCS was considered. In the limiting heat addition transient, a maximum peak pressure of 418 psig was calculated.
'{ta. esti"C tvssE W&
o"c= has identified in letter PSE-93-204 dated March 15, 1993 (NSAL-93-005B) a non-conservatism in the calculation for peak pressure for the heat input and mass addition transients that affects both Salem Units ] and 2. The cencern is that the difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg pressure and the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature (P/T) limits are defined) with the reactor coolant pumps (RCP) operating was not considered in the original Salem POPS analysis. This~results in encroachment on the P/T limits. To quantify the effects on Salem, specific pressure differences associated with RCP operation have been calculated for one, two and four RCPs operating. The results of these calcu)ations pro *rided values of 29, 37, and 71 psig with one, two and four RCPs operating, respectively. A correction pressure of 3.0 psig has been added to these values to account for transmitter elevation differences not previously accounted for in the calculations. As a result, procedure revisions have been implemented to limit the number of RCPs in operation to 1 while in Mode 5 (< 200*F). In addition, plant operating procedures require that the power supplies be removed from both SI pumps (675 GPM) upon entry into Mode 4 (< 350*F). Therefore, only the mass input from a Centrifugal Charging Pump (CCP) (560 GPM) is considered. The net results, assuming operation of a CCP and the pressure difference from operation of one RCP, are peak pressures below the specified limits in the Unit 1 and 2 P/T curves. Since the completion of these evaluations, it has now been determined that the Positive Displacement Charging Pump (PDP), if already in operation, would continue to operate upon initiation of a SI signal if offsite power remains available. During this postulated event, letdown would be automatically isolated as part of the SI actuation. The additional flow from the PDP is a concern for the period of time when the RCS is < 200*F (Mode 5), the PDP is in operation, and one (1) CCP has its associated power supply available. Therefore, the combined flow of 665 GPM from the PDP (105 GPM) and the CCP (560 GPM) is now considered the most limiting mass addition transient.
e PSE&G has re-analyzed the mass addition event using the GOTHIC computer code assuming a bounding maximum combined pump flow rate of 675 gpm. 'The resulting peak pressure is 443 psig. After including the pressure differential due to the operation of one l RCP and the 2 psig elevation correction, a peak pressure of 474 psig is established. This pressure exceeds the design basis limit of 450 psig for Salem Unit 1. The P/T limit of 475 psig for Unit 2 continues to be met with relatively no margin. These results are summarized in the follow.ng table: UNIT POPS CALC. PEAK AP W/ 1 RCP CORRECTED T/S SETPOINT PRESS. RUNNING PEAK RCS PRESS. (PSIG) (PSIG) & ELEV. PRESSUT.E LIMITS - BASED ON CORRECTION (PSIG) HEATUP OR MASS INPUT (PSIG) 20*T/HR CASE W/ 675 CoOLDOWN GPM FLOW (PSIG) S1 375 443 31 474 450 S2 375 443 31 474 475 For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature), the original peak pressure calculated was 410 psig. When the pressure differential due to the operation of one RCP (29 psi) and the 2.0 psig to correct for elevation is added, the peak pressure is 449 psig. This peak pressure remains below the POPS limits of 450 and.475 psig for Salem Units 1 and 2, respectively. As compensatory action, administrative controls ensure RHR relief valve RH3 is available and the associated RHR isolation valves are in the open position. This action is necessary when RCS temperature is less than 200 degrees F, the PDP is in operation and a CCP has power available. Analyses for a mass addition transient up to 780 GPM, assuming either 2 PORVs, or 1 PORV and RH3 are available, have determined that sufficient relieving capacity exists to ensure that the current P/T limits will not be exceeded. This is not required if a vent path is established instead of POPS in accordance with Tech. Specs., or the PDP is not in operation. Salem Unit 1 operating procedures limit operation in Mode 5 to one RCP, require that power be removed from the SI pumps in Mode 4, and require relief valve RH3 to be available to ensure the current P/T limits are met when RCS temperature is < 200*F. Similar administrative controls for Salem Unit 2 are in place although RH3 is not needed to meet its P/T limits. However, without credit for'the relieving capa, city of RH3, relatively no e re- '
I opera'.ing margin between the current P/T limits calculated in accordance with 10CFR50, Appendix G and the peak pressure during , low temperature overpressure transients exists for Salem Unit 2. I APPARENT CAUSE OF OCCURRENCE: I
, a% b Je - l Basedontheresulsok,theinvestigationto-date, the operation I
( of the PDP was no considered in the original POPS analysis nor was it considered in any other Chapter 15 accident analyses. It ; was believed that the PDP operation would be terminated following initiation of a Safety Injection (SI) signal. Following the G April 7 event at Salem Unit 1, it was determined that the PDP
, would continue to operate if already in service following initiation of a SI signal with offsite power available. This was confirmed by a review of the current gentrol configuration for the PDP. Tk cmuh q ,.t..J % +o ~ . 7 N. W-t niG.a.,m~ I n ff 4 L 4 6 g 6- 4.
r,su m ,LLL LJ ..< x ~ t W % & Ws ..n , 6.L k ..'n Following the April 7 event, the NSSS Vendor completed an sl -] 4 u r
\x evaltation of the continued PDP operation during SI for the '
complete spectrum of Chapter 15 accidents including the Inadvertent SI at Power transient. The evaluation also addressed
. the impact of continued PDP operation on affected systems and component performance. The NSSS Vendor evaluation concluded that continued operation of the PDP would not impact existing accident analyses. However, this evaluation did not consider the PDP operation during low temperature conditions.
PRIOR SIMILAR OCCURRENCES: No other prior similar occurrences have been identified related to this design deficiency. / SAFETY SIGNIFICANCE: , 1 This event is reportable in accordance with the requirements of I 10CFR50. 73 (a) (2) (ii) (B) , due to the POPS not being able to meet f it design basis, and provide adequate overpressure protection should a mass aadition transient occur. 1% L N 4 W a nes+- ' f> * '*.1 k Hy u.~% .!,+ . weap, '"1 \ tow A 4 Additional margin on the Tech. Spec. curves can be gained when I operating with POPS (RCS cold legs less than 312*F) by taking I credit for ASME Code Case N-514. This Code Case states that the \ ' LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of ASME Section XI, Article G-2215. Crediting the Code Case will allow the maximum allowable pressure (Tech. Spec. P/T limits) for POPS to be increased to 495 psig and 522.5 psig for Salem Units 1 and 2, respectively. The calculated peak pressure of 474 psig would be considerably below the P/T limits assuming the additional 10% / margin allowed by Code Case N-514. (It is noted that the I contents of Code Case N-514 have been incorporated into Appendix . G of Section XI of the ASME Code in the 1993 Addenda to Section XI.) I\oHim s* 4 M'ob
- c a I.
/
The ASME Working Group on Operating Plant Criteria (WGOPC) developed code guidelines to define POPS limits that will avoid certain unnecessary operational restrictions, provide adequate margins against failures, and reduce the poiential for unnecessary activation of pressure relieving devices used for POPS (PORVe). The philosophy used by the WGOPC-for developing these guidelines was to ensure that the POPS limits are still below the P/T limits during normal plant operation, but allows the pressure that may occur with activation of the POPS to exceed the P/T limits by a maximum of 10% provided acceptable margins are maintained during these events. This philosophy was to protect the reactor vessel from low temperature overpressure transients and still maintain the P/T limits in the Technical Specifications applicable for normal heatup and cooldown in accordance with Appendices G to 10CFR50 and Section XI of the l ASPE Code. The WGOPC applied deterministic and probabilistic analysis techniques for several different flaw locations and heat-up and l cooldown rates to establish the conditions delineated by the Code Case. For consideration, there are several conservatisms inherent in the development of the ASME Section XI, Appendix G P/T curve calculations that include:
- 1) The safety factor of 2 on the principal membrane (pressure) stresses.
- 2) A margin factor applied to RT in accordance with the requirementsofRegulatoryGu$be1.99, Revision 2 (eg., 2-sigma margins are applied in determining the adjusted reference temperature).
- 3) The disregarding of increased mechanical properties of the reactor vessel which accompany material embrittlement (elevated yield strength and flow stress) .
- 4) An assumed flaw in the wall of the reactor vessel that has a depth equal to 1/4 the thickness of the vessel wall and a length equal to 1-1/2 times the vessel wall thickness.
- 5) The reference stress intensity curves defined in Section XI, Appendix G are used to bound the dynamic crack initiation and crack arrest toughness.-
1
- i 1 i i
l In summary, Code Case N-514 allows exceedance of the P/T limits calculated in_accordance with 10CFR50, Appendix 0 by 10%. The ' application of Code Case N-514 provides sufficient margin such i that-the inadvertent SI actuation that_results in a mass input from both a CCP and'the PDP will not result in any P/T limits ; } 'being exceeded assuming the availability of only one POPS. Code
; -Case N-514 also allows the operation of up to two RCPs when RCS
- temperature is < 200*F, and removes the Salem _ Unit i requirement ,
4 for RH3.- Therefore, it is concluded that this event-does not : have any. safety significance. CORRECTIVE ACTIONS: i Salem Unit 1 operating procedures limit operation in Mode 5 to one RCP, and require that power be removed from the SI pumps in .
- Mode 4. Relief valve RH3 is required to be available to ensure ?
E the current P/T limits are met when RCS temperature is < 200*F ,
- when the PDP is in operation. Relief- valve RH3 is not required l l to be available if a vent path is established instead of POPS in i accordance with Tech. Specs.,
1 Similar. administrative controle for Salem Unit 2 are in place 4 although RH3 is not needed to meet its P/T limits. A letter to the NRC is currently under preparation requesting Lapplication of Code Case N-514 which allows exceedance of the Pressure / Temperature (P/T) limits _ calculated in accordance with the requirements of 10CFR50, Appendix G by 10%. This additional' I 10% margin will increase the P/T limits operating margin for the POPS ' during I,ow Temperature Overpressure Protection (LTOP)
- conditions, and allow the operating restrictions discussed above to be removed.
'A. License Change Request is_also under consideration to credit '
, RH,?-as part of POPS. 1 PSE&G is also evaluating the benefits of' '
- completing a design change to terminate PDP operation following-initiation of a SI signal if offsite power remains available.
The NSSS. Vendor has' completed-an evaluation of the continued PDP ' operation during SI for the complete spectrum of Chapter 15 i accidents including the Inadvertent SI at Power transient. The evaluation also addressed the impact of continued-PDP operation on affected systems and component performance. The NSSS Vendor i evaluation concluded that continued operation offthe PDP would U not. impact existirg accident analyses. Ech J A ' -l C p s ) i
$m a u a a Q, u :q ' ' )
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+ DESCRIPTION OF OCCURRENCE OnL November 17, 1994, Salem Unit l'was_in Mode.1 at 100% power. !
10 n that_ day, PSE&G determined that the. pressure limit of 450 psig ; may'be exceeded for Salem Unit 1 when considering the additional j flow - f rom .the? Positive Displacement Pump _ (PDP) . The Technical: ; Specification Pressure / Temperature limits for plant.heatup ! (Figure 3.4-2)1and~cooldown (Figure 3.4-3), which are determinod-in accordance with 10CFR50, Appendix G, ensure reactor vessel ' , l integrity.- The current _ bases for the Pressurizer: Overpressure. Protection System (Technical _ Specification 3/4.4.9.3) states'that j one : POPS : relief valve ' provides adequate relieving capacity in the event'of an overpressure transient that includes.the inadvertent- ' start of a SI' pump-(Mass Addition transient) into a water' solid- i Reactor Coolant System (RCS). PSE&G has determined that the; following< realistic mass _ addition _ transient assumptions when
-considering the-additional flow'from the PDP wodid_ place Salem i Unit 1 outside the limiting design and licensing basis POPS '
.i analysis:
l
- RCS temperature less than1200*F
- One Reactor _ Coolant Pump.(RCP) in operation
< PDP in service *
- a maximum of One' Centrifugal. Charging Pump (CCP)s ha3. j ts power supply available {
Initiat' ion of a SI. signal _would result;in a mass' addition from a combination of-flow from the'PDPLand.the CCP based on the-above' ; assumptions, and could-result in a peak RCS pressure of 474 psig , that exceeds the design basis pressure limit of-450 psig for 1 Salem Unit 1.- , ANALYSIS OF OCCURRENCE The-Pressurizer Overpressure Protection. System:(POPS) protects-the Reactor' Coolant' System (RCS) from exceeding the Tech. Spec.
-Pressure / Temperature.(P/T)-limits for. plant heatup (Figure.3.4-2) cand cooldown (Figure 3.4-3) by opening.the Power Operated Relief Valves (PORV) during cold overpressure transients (RCS cold 11eg ,
temperature.below 312*F). The limits are determined-in accordance with the vacuirements of 10CFR50, Appendix G. .
- According to the exi. ng design bases, either-PORV:has adequate ,
-relieving capacity to , otect the RCS fr m overpressurization ;
when the transient is limited to either~(1) the start of an idle
~
RCP.with the secondary water temperature less than.or equal'to 50*F above the RCS cold leg temperature (heat' addition), or (2) the start.of a Safety Injection-(SI) pump _and_ resultant injection Jinto'a water solid RCS (mass addition). 01 . vy--- isyy.i- --p 9- YTW---' 7 gr M-1 y--a-ig--T--wv-
i V The pressure limits at the low temperature end of the P/T curves are 450 and 475 psig for Salem Units 1 and 2, respectively, as read from the current heatup and cooldown curves (Tech. Spec. Figures 3.4-2 and 3.4-3, respectively). The original Salem POPS analysis calculated-a maximum peak pressura for the_most limiting mass addition transient of.446 psig with the PORV set at a pressure of 375 psig. In this analysis, the RCS pressure due to injection of 780 gpm. flow into an initially cold water solid RCS was considered. In the limiting heat adcition transient, a maximum peak pressure of 418'psig was calculated. The NSSS vendor identified.in letter PSE-93-204 dated March 15, ;
-1993 (NSAL-93-005B) a non-conservatism in the calculation for j peak pressure for the heat input and mass addition transients t that affects beth Salem Units 1 and 2. The concern is that the !
difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg pressure and the , reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature (P/T) limits are defined) with the l reactor coolant pumps (RCP) operating was not considered in the original Salem POPS analysis. This recults in encroachment on the P/T limits. To quantify the effects on Salem, specific pressure differences associated with RCP operation have been calculated for one, two-and four RCPs operating. The results of these calculations ' provided values of 29, 37, and-71 psig with one, two and four r RCPs operating, respectively. A correction pressure of 2.0 psig-has been_added to these values to account:for transmitter , elevation differences not previously accounted for in the calculations. As a result, procedure revisions have been implemented to limit the number of RCPs in operation to 1 while j in Mode 5 (< 200*F). In addition, plant: operating procedures j I require that the power supplies.be removed from both1sI pumps '
-(675 GPM) upon entry into Mode 4- (< ' 3 50
- F) , Therefore,-only the 1 mass input from a Centrifugal Charging Pump :(CCP) (560 GPM) is considered. The net results, assuming operation of a CCP and the
' pressure difference from operation of one RCP, are peak pressures i below the specified limits in the Unit 1 and 2 P/T curves.
Since the_ completion of these evaluations, it has now been determined that_the Positive Displacement-Charging Pump (PDP), if i Lalready in operation, would continuo to operate upon initiation ofca SI signal if offsite power remains available. During this postulated event, letdown would be automatically isolated as part of the SI. actuation. The additional flow from-the PDP-is a-concern for the period of time when the RCS is < 200'F (Mode 5), the'PDP is in operation, and one (1) -CCP: has its associated- power supplyiavailable. .Therefore,_the combined flow of 665-GPM frcm
-the-PDP-(105 GPM)'and-the CCP (560 GPM) is now considered the most limiting mass addition transient.. -
I PSE&G has re-analyzed the mass addition event using the GOTHIC computer code assuming a bounding maximum combined pump flow rate of 675 gpm. The resulting peak pressure is 443 psig. After including the pressure differential due to the operation of one RCP and the 2 psig elevation correction, a peak pressure of 474 psig is established. This pressure exceeds the design basis limit of 450 psig for Salem Unit 1. The P/T limit of 475 psig for Unit 2 continues to be met with relatively no margin. .These results are summarized in the following table: 1
\
! UNIT POPS CALC. PEAX aP W/ 1 RCP CORRECTED T/S SETPOINT PRESS. RLHNING PEAK RCS PRESS. (PSIG) (PSIG) t. ELEV. PRESSURE LIMITS - BASED ON CORRECTION (PSIG) HEATUP OR MASS INFUT (PSIG) 20*F/HR CASE W/ 675 c00LDOhH GPM FLOW (PSIG) S1 375 443 31 474 450 S2 375 443 31 474 475 For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature), the original peak pressure calculated was 418 psig. Wnen the pressure differential due to the operation of one RCP (29 psi) and the 2.0 psig to correct for elevation is added, the peak pressure is 449 psig. Tuis peak pressure remains below the POPS limits of 450 and 475 poig for Salem Units 1 and 2, respectively. As compensatory action, administrative controls ensure RHR relief valve RH3 is available and the associated RHR isolation valves are in the open position. This. action is necessary when RCS temperature is less than 200 degrees F, the PDP is in operation and a CCP has pcwer available. Analyses for a mass addition transient up to 780 GPM, assuming either 2 PORVs, or 1 PORV and RH3 are available, have determined that sufficient relieving capacity exists to ensure that the current P/T limits will not be exceeded. This is not required if a vent path is established instead of POPS in accordance with iach. Specs., or the PDP is not in operation. Salem Unit 1 operating procedure: limit operation in Mode ~5 to one RCP, require tha*. power be removed from the SI pumps in Mode 4, and require relief valve RH3 to be available to ensure the current P/T limits are met when RCS temperature is < 200*F. Similar administrative controls for Salem Unit 2 are in place although RH3 is not needed to meet its P/T limits. However, without credit for the relieving capacity of RH3, relatively no
4 operacing margin between the current T/*2 linsr.2 calculated in accordance with 10CFR50, Appendix G ar.6 rAe Teat pressure during low temperature overpressure transients exists f>or Salem Unit 2. APPARENT CAUSE OF OCCURRENCE: Based on the results of the investigation to-date, the operation of the PDP was not considered in the original POPS analysis nor was 4,t considered by the NSSS Vendor in any other Chapter 15
. accident analyses. It was believed that the JDP operation would be terminated following initiation of a Safety Injection (SI) signal. Following the-April 7 event at Salem Unit 1, it was l determined that the PDP would continue to operate if already in l
' service following initiation of a SI signal with offsite power available. This was confirmed by a review of the current control configuration for the PDP. The cause of this omission has yet to be determined at this time, l The initial POPS analysis for the mass addition transient also did not account for the differential pressure between the mid-plane of the core and the location of the pressure sensors located in the RCS hot legs with the RCP(s) in operation. Following the April 7 event, the NSSS Vendor completed an evaluation of the continued PDP operation during SI for the complete spectrum of Chapter 15 accidents including the Inadvertent SI at Power transient. The evaluation also addressed the impact of continued PDP operation on affected systems and component performance. The NSSS Vendor evaluation concluded that continued operation of the PDP would not impact existing accident analyses. However, this evaluation did not consider the PDP cperation during low temperature conditions. PRIOR SIMILAR OCCURRENCES: No other prior similar occurrences have been identified related to this design deficiency. SAFETY SIGNIFICANCE: This event is reportable in accordance with the requirements of 10CFR50. 73 (a) (2) (ii) (B) , due to the POPS not being able to meet it. design basis, and provide adequate overpressure protection should a mass addition transient occur. This LER also satisfies any reporting requirements pursuant to 10CFR21. Additional margin on the Tech. Spec. curves can be gained when operating with POPS (RCS cold legs less than 312*F) oy taking credit.for ASME Code Case N-514. This Code Case states that the LTOP systems shall limit the maximum pressure in the vessel to 110% of the-pressure determined to satisfy Appendix G of ASME. Section XI, Article G-2215. Crediting the Code Case will allow
4 the maximum allowable pressure (Tech. Spec. P/T limits) for POPS to be increased to 495 psig and 522.5 psig for Salem Units 1 and 2, respectively. The calculated peak pressure of 474 psig would be considerably below the P/T limits assuming the additional 10% margin allowed by Code Case N-514. (It is noted that the contents of Code Case N-514 have_been incorporated into Appendix G of Section XI of the ASME Code in the 1993 Addenda to Section XI.) i The ASME Working Group on Operating Plant Criteria (WGOPC) l developed code guidelines to define POPS limits that will avoid i certain unnecessary operational restrictions, provide adequate margins against failures, and reduce the potential for i unnecessary activation of pressure relieving devices used for i POPS (PORVs). The philosophy used by the WGOPC for developing these guidelines was to ensure that the POPS limits are still below the P/T limits during normal plant operation, but allows the pressure _that may. occur with activetion of the POPS to exceed the P/T limits by a maximum of 10% provided acceptable margins are maintained during these events. This philosophy was to , protect the reactor vessel from low temperature overpressure ' transients and still maintain the P/T limits in the Technical
-Specifications applicable for normal heatup and cooldown in accordance with Appendices G to 10CFR50 and Section XI of the ASME Code.
The WGOPC applied deterministic and probabilistic analysis techniques for several different flaw locations and heat-up and cooldown rates to establish the conditions delineated by the Code Case. T For consideration, there are several conservatisms inherent in the development of the ASME Section XI, Appendix G P/T curve calculations that include:
- 1) The safety factor of 2 on the principal membrane (pressure) stresses.
- 2) A-margin factor applied to RTyp, in accordance with the requirements of Regulatory Guide 1.99, Revision 2 (eg., 2-s13ma margins are applied in determining the adjusted retirence temperature).
- 3) The disregarding of increased mechanical properties of the reactor vessel which accompany material embrittlement (elevated yield strength and flow stress).
- 4) An assumed flaw in the wall of the reactor vessel that has
- a. depth equal to.1/4 the thickness of the vessel wall and a length equal to 1-1/2 times the vessel wall thickness.
l-
6
.5) The reference stress intensity curves defined in Section XI, Appendix-G-are used to bound the dynamic crack initiation and crack arrest toughness.
In' summary, Code. case N 514 allows exceedance-of the P/T limits calculatedrin accordance with 10CFR50, Appendix G by 10%. The
- application cf Code Case N-514-provj is sufficient margin-such that the inadvertent SI actuation that results in a mass input-
)
from both a CCP and the PDP will not result in any P/T limits J being exceeded assuming the availability of only one POPS. Code !
-Case N-514-also allows the operation of up to-two RCPs.when RCS- )
temperature-is < 200'F, and removes the Salem Unit I requirement for RH3. Thereforo, it is concluded that this event does not have any safety significance. CORRECTIVE ACTIONS: Salem Unit 1 operating procedures limit operation in Mode 5 to one RCP, and require that power be removed from the SI pumps in
' Mode 4. Relief valve lRH3 is required-to be available to ensure the current P/T limits are met when RCS temperature is < 200*F when the PDP is in operation. Relief valve RH3 is not required to be available if a vent path is established instead of POPS in accordance with Tech. Specs..
Similar administrative controls for Salem Unit 2 are in place although RH3 is not needed to meet its P/T limits. A letter to the NRC is currently under preparation equesting application of-Code Case N-514 which allows exceedance of the H Pressure / Temperature (P/T) limits calculated in accordance with ! the--requirements of 10CFR50, Appendix G by 10F. This additional 10% margin will increase the P/T. limits operating margin for the POPS during Low Temperature Overpressure Protection (LTOP) _ conditions, and allow the operating restrictions discussed above to be removed.
!A1 License Change Request _is also under consideration to credit RH3 as part of POPS. PSELG is also evaluating the benefits of completing a design :hange to terminate PDP operation following initiation of a--SI signal if'offsite power remains available.
The.NSSS Vendor has completed an evaluation of the-continued PDP
' operation during SI'for the complete spectrum of. Chapter 15~
accidents including the Inadvertent SI at Power transient. The-evaluation also addressed the impact of continued PDP operation on affected systems and component-performance. The NSSS Vendor evaluation concluded that continued operation of-the PDP would not11mpact existing accident analyses. PSE&G has requested the NSSS Vendor to investigate any-generic implications associated' p with the continued operation of the PDP during Safety Injection. N L l l:
F t' h *j DESCRIPTION OF-OCCURRENCE On November 17, 1994,1 Salem Unit 1 was in Mode 1 at 100%. power. . On.that day, PSE&G determined that the pressure limit of 450 psig may be exceeded.for Salem Unit.1 when considering the additional ! flow lfrom the Positive Displacement Pump (PDP)_. The_ Technical
. Specification Pressure / Temperature limits for plant heatup j (Figure 3.4-2) and cooldown1(Figure 3.4-3),7which are determined ;
'in accordance with 10CFR50, Appendix G, ensure reactor vessel -{
integrity. The. current bases for the Pressurizer Overpressure i Protection System (Technical Specification 3/4.4.9.3) states that
-one POPS relief valve provides adequate relieving capacity,in the event of an overpressure transient that includes the inadvertent !
start of a SI pump-(Mass Addition transient) into a water solid' !
' Reactor Coolant System (RCS). PSE&G has_ determined that the !
- following realistic mass addition transient assumptions when- .;
considering thefadditional flow;from the PDP would place Salem ^
. Unit 11 outside the design and licensing basis POPS analysis: .!
P
- RCS temperature less than 200'F
- One Reactor Coolant Pump (RCP) in operation PDP in service "
- a ; maximum of One Centrifugal charging Pump - (CCP) has its power
. supply available Initiation of a SI signal would result:in a mass addition'from a. i combination of flow from the PDP and the CCP based on the above -
' assumptions; and could result'in a peakJRCS pressure of 474 psig:
that exceeds the design basis pressure limit of 450 psig for Salem Unit 1. ANALYSIS OF OCCURRENCE' Backoround i
-The Pressurizer Overpressure Protection System (POPS). protects the. Reactor Coolant System (RCS) from exceeding the Tech, Spec._ j Pressure / Temperature f(P/T)' l'imits for plant heatup (Figure 3.4-2) and cooldown-(Figure 3.4-3) by opening the Power Operated Relief Valves .(PORV) during cold overpressure transients (RCS cold leg temperature below 312'F).: The limits are determined in '
accordance with the requirements of 10CFR50, Appendix G.
. According1to the' existing design bases, either PORV has adequate . ' , relieving capacity to' protect.the:RCS from overpressurization when=the. transient is limited to either (1) the start of an idle RCP with_the secondary water temperature less than or equal-to L50'F above the RCS cold leg temperature'(heat addition), or -(2) -
thelstart of a. Safety Injection (SI) pump and resultant injection 4 into a water! solid 1RCS (mass addition). -- 1 L
+p.
w , 'w>- ,m + - m p m - w-- ,-,---n-,-e- ~ s ww..-e,.mn.u--a w aw, --w, w--~mwn -.v ~e =m- -tw- ,
i ) The pressure limits at the low temperature end of the P/T curves are 450 and 475 psig for Salem Units 1 and 2, respectively, as read from the current heatup and cooldown curves (Tech. Spec. Figures 3.4-2 and 3.4-3, respectively). The original POPS ! analysis calculated a maximum peak pressure for the most limiting mass addition transient of 446 psig with the PORY set at a pressure of 375 psig. In this analysis, the RCS pressure due to safety injection (780 gpm) into an initially cold water solid RCS was considered. In the limit.ing heat addition transient, a maximum peak pressure of 418 neig was calculated. The NSSS vendor identified in letter PSE-93-204 dated March 15, 1993 (NSAL-93-005B) a non-conservatism in the calculation for peak pressure for the heat input and mass addition transients that affects both Salem Units 1 and 2. The identified concern was that the difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg pressure and the reactor vessel midplane (where the Tech. Spec, heatup and cooldown pressure / temperature (P/T) limits are defined) with the reactor coolant pumps (RCP) operating was not considered in the original POPS analysis resulting in encroachment on the P/T limits. To quantify the effects on Salem, specific pressure differences associated with RCP operation were calculated for one, two and four RCPs operating. The results of these calculations provided values of 29, 37, and 71 psig with one, two and four RCPs operating, respectively. A correcticn pressure of 2.0 psig , was then added to account for transmitter elevation differences not previously accounted for in the calculations. Existing plant operating procedures require that the power supplies be removed from both SI pumps (675 GPM) upon entry into Mode 4 (< 350*F). Procedure revisions were implemented to limit the number of RCPs in operation to 1 while in Mode 5 (< 200
- F) . Therefore, only the mass input from a Centrifugal Charging Pump (CCP) (560 GPM) needed to be considered. The net results were peak pressures below the specified limits in the Unit 1 and 2 P/T curves.
Present situation Since the completion of the NSAL evaluations, it has now been determined that the Positive Displacement Charging Pump (PDP), if already in operation, would continue to operate upon initiation of a SI signal if offsite power remained available. During this postulated event, letdown would. automatically isolate as part of the SI actuation.. The additional flow from the PDP is a concern for the limited period of time when the RCS is < 200*F (Mode 5), the PDP is in operation, and one (1) CCP has its associated power supply available. The combined flow of 665 GPM from the PDP (105 GPM) and the CCP (560 GPM) is now considered the most limiting mass addition transient,
E
?
PSE&G has re-analyzed the mass addition event using the GOTHIC computer code assuming a bounding maximum combined pump flow rate of 675 gpm. The resulting peak pressure is 4t3 psig. After including the original NSAL concerns,-a peak pressure of 474 psig is established. This pressure exceeds the current limit of 450 psig for Salem Unit 1. The P/T limit of 475 psig for Unit 2 continues to be met with relatively no margin. These results are summarized in the following table: UNIT POPS CALC. PEAK AP W/ 1 RCP CORRECfED T/S SETPOINT' PRESS. RUNNING PEAK RCS PRESS. (PSIG) (PSIG) & ELEV. PRES $URE LIMITS - BASED ON CORRECTION (PSIG) HEATUP OR MASS INPUT (pr a) 20'F/HR CASE W/ 675 COOLDOWN GPM FLOW (PSIG) S1 375 443 31 474 450 S2 375 443 31 474 475 _ For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature less than or equal to 50'F above the RCS cold leg temperature), the original peak pressure calculated was 418 psig. When the pressure differential due to the operation of one RCP (29 psi) and the 2.0 psig to correct for elevation is added, the peak pressure le 449 psig. This peak pressure remains below the POPS limits of 450 and 475 psig for Salem Units 1 and 2, respectively. As ccmpensatory action, administrative controls ensure RHR relief valve RH3 is available and the associated RHR isolation valves are in the open position._ This action is necessary when RCS temperature is less than 200 degrees F, the PDP is in operation and a CCP has power available. Analyses for a mass addition transient up to 780 GPM, assumine either 2 PORVs, or 1 PORV and RH3 are available, have determined that sufficient relieving capacity exists to ensure that the current P/T limits will not be exceeded. APPARENT CAUSE OF OCCURRENCE: Based on the results of the investigation tv-date, the operation of the PDP was not considered in the original POPS analysis nor was it considered by the NSSS Vendor in any other Chapter 15 accident analyses. It was believed that the PDP operation would be terminated following initiation of a Safety Injection 'dI) signal. Following the April 7 event at Salem Unit 1, it was determined that the PDP would continue to operate if already in service following initiation of a SI signal with offeite power
avails le. This was confirmed by a. review of the current control , configuration'for the PDP. The cause of this omission has yet to be determined at this. time. The NSSS Vendor has been requested to investigate any generic implications associated with the i continued operation of the PDP during Safety' Injection (eg. . ! reportability in accordance with 10CFR21). ! The-initial POPS analysis foi the mass = addition transient also did not account for the differential pressure between the mid- ,
. plane of the core and the location of the pressure sensors '
located in the RCS hot legs with the RCP(s) in operation . Following the April 7 event, the NSSS Vendor' completed an i evaluation of continued PDP operation during SI for the complete l spectrum'of_ Chapter 15 accidents including the Inadvertent SI at -; Power transient. The evaluation also addressed the impact of continued PDP operation on-affected systems and component performance. The NSSS Vendor evaluation concluded that continued ! operation of the PDP would not impact' existing accident analyseo. fHowever, this evaluation did not consider the PDP operation. l' during low temperature conditions. PRIOR SIMILAR OCCURRENCES: No other prior similar occurrences have been identified related to this design deficiency. SAFETY SIGNIFICANCEr This event is reportable in accordance with the requirements of
- 10CFR50. 73 (a)'(2) (ii) (B) , due to the POPS not being able to meet. '
' it current design basis, and provide adequate overpressure protection should a mass addition transient occur. This LER also ;
satisfies =any' reporting requirements pursuant;to 10CFR21.
~
~
Additional margin- on the' Tech. Spec. curves can be gained when-operating'with: POPS (RCS cold legs less than 312*F) by taking credit for ASME Code-' Case'N-514. This Code Case states that the LTOP systemsLshall limit the maximum pressure in the vessel to- i 110% of the1 pressure determined to satisfy Appendix'G of ASME
.Section.XI, Article G-2215. Crediting the Code Case will allow :
the maximumLallowable pressure'(Tech.. Spec. P/T limits) for POPS
- to be increased to 495 psig'and 522.5 psig for Salem Units 1 and 2~, respectively. - The calculated peak. pressure of 474 psig would be considerably below the P/T limits assuming the additional 10%
4 sesm -- e'e .4 ., ., eE _ ,6 - .--w,_w.,-c.
~ w-- ,vi-y.-m.-- , - ,wn g , ,- -- -~ w n-- e m- ,--+7r---- .%---
e margin allowed by Code Case N-514. (It is noted that the contents of Code Case N-514 have been incorporated into Appendix G of Section XI of the ASME Code in the 1993 Addenda to Section XI.) The ASME Working Group on Operating Plant Criteria (WGOPC) developed code guidelines to define POPS limits that will avoid certain unnecessary operational restrictions, provide adequate margins against failures, and reduce the potential for unnecessary activation of pressure relieving devices used for POPS (PORVs). The philosophy used by the WGOPC for developing these guidelines was to ensure that Lne POPS limits are still below the P/T limits during norma' plant operation, but allows the pressure that may occur with activation of the POPS to exceed the P/T limits by a maximum of 10% provided acceptable margins are maintained during these events. This philosophy was to protect the reactor vessel from low temperature overpressure transients and still maintain the P/T limits in the Technical Specifications applicable for normal heatup and cooldown in accordance with Appendices G to 10CFR50 and Section XI of the ASME Code. The WGOPC applied deterministic and probabilistic analysis techniques for several different flew locations and heat-up and ccoldown rates to establish the conditions delineated by the code Case. For consideration, there are several conservatisms inherent in the development of the ASME Section XI, Appendix G P/T curve calculations that include:
- 1) The safety factor of 2 on the principal membrane (pressure) stresses.
- 2) A margin factor applied to RT in accordance with the requirementsofRegulatoryGu!$e1.99, Revision 2 (eg., 2-sigma margins are applied in determining the adjusted reference temperature).
- 3) The disregarding of increased mechanical properties of the reactor vessel which accompany materia t embrit' 2 ment (elevated yield strength and flow stresa).
- 4) An assumed flaw in the wall of the reactor vessel that has a depth equal to 1/4 the thickness of the vessel wall and a length equal to 1-1/2 times the vessel wall thickness.
- 5) The reference stress intensity curves defined in Section XI, Appendix G are used to bound the dynamic crack initiation and crack arrest toughness.
O i . 1 In summary, code Case N-514 allows exceedance of the P/T limits calculated-in accordance with 10CFR50, Appendix G by 10%. The application of Code Case N-514 provides sufficient margin such j that the inadvertent SI actuation that results in a mass input from both a CCP.and the PDP will not result in any P/T limits i being exceeded assuming the availability of-only one POPS. Code ' Case N-514.also allows the operation of up to two RCPs when RCS temperature is < 200*F, and removes the Salem Unit i requirement for RH3. Therefore, it is concluded that this event does not
- have any safety significance. ,
~ CORRECTIVE ACTIONSr Salem Unit 1 operating procedures-limit operation in Mode 5 to one PCP, and require that power.be removed from the SI pumps in :
Mode 4. Relief valve RH3 is required to be available to ensure the current P/T limits are met when RCS temperature is <.200*F
-when the PDP is in operation. Relief valve RH3 is not required to be available if a vent path is established instead of POPS in
.accordance with Tech. Specs..
i similar administrative controls for. Salem Unit 2 are in place '
- although RH311s not needed to meet its P/T limits.
A letter-to the NRC is currently under. preparation requesting applicationfof= Code Case N-514 which allows exceedance of the Pressure / Temperature (P/T) limits calculated in accordance with the requirements of 10CFR50, Appendix G by 10%. This additional 10% margin will increase the P/T limits operating margin for the POP 3 during Low Temperature Overpressure. Protection (LTOP) conditions, and allow the oparating restrictions discussed above to be removed. A License Change Request.is also under consideration to credit t RH3 as part.of POPS. .PSE&G is_also evaluating the benefits ~of-completing a design change to terminate PDP operation following initiation of a SI signal if offsite power remains available, i The NSSS Vendor has completed an evaluation of the continued PDP
- operation during SI for the complete spectrum of Chapter 15 i: . accidents including the Inadvertent SI at Power-transient. The evaluation also addressed the impact of-continued PDP operation
- on.affected systems and component performance. The NSSS Vendor ! -
- . evaluation concluded that continued operation of the PDP would not' impact existing-accident analyses. PSE&G.has requested.the NSSS Vendor to investigate any generic implications associated with the continued operation of the PDP during Safety Injection,i 1
k
- , - . . .r. . , , _.- <, .,r,. , , , , . , . . - .,.
s D l ), l 1 DESCRIPTION OF OCCURRENCE: On November 17, 1994, Salem Unit 1 was in Mode 1 at 100% power. On that daye PSE&G determined that the pressure limit of 450 psig may be exceeded for Salem Unit 1 when considering the additional flow from the Positive Displacement Pump (PDP). The Technical Specification Pressure / Temperature limits for plant heatup (Figure 3.4-2) and cooldown (Figure 3.4-3), which are determined in accordance with 10CFR50, Appendix G, ensuro reactor vessel integrity. The current bases for the Pressurizer Overpressure Protec* ion System (Technical Specification 3/4.4.9.3) states that one POPS relief valve provides adequate relieving capacity in the event of an overpressure transient that includes the inadvertent start of a SI pump (Mass Addition transient) into a water solid Reactor Coolant System (RCS). PSE&G has determined that the following realistic mass addition transient assumptions when considering the additional flow from the PDP would place Salem Unit 1 outside the limiting design and licensing basis POPS analysis: RCS temperature less than 200'F
- One Reactor Coolant Pump (RCP) in operation
- PDP in service a maximum of One Centrifugal Charging Pump (CCP) has its power supply avai]able Initiation of a SI signal would result in a mass addition from a combination of flow from the PDP and the CCP based on the above assumptions, and could result in a peak RCS pressure of 474 psig that exceeds the design basis pressure limit of 450 psig for Salem Unit 1.
ANALYS'S OF OCCURRENCE The Pressuricer Overpressure Protection System (POPS) protects the Reactor Coolant System (RCS) from exceeding the Tech. Spec. Pressure / Temperature (P/T) limits for plant heatup (Figure 3.4-2) and cooldown (Figure 3.4-3) by opening the Power Operated Relief Valves (PORV) during cold overpressure transients (RCS cold leg temperature below 312'F) . The limits are determined in accordance with the requirements of 10CFR50, Appendix G. According to the existing design bases, either PORV has adequate relieving capacity to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature (heat addition), or (2) the start of a safety Injection (SI) pump and resultant injection into a water solid RCS (mass addition). s
+
/' , The pressure limits at the low temperature end of the P/T , l are 450 and 475 psig for Salem Units 1 and 2, respectiv' read from the current heatup and cooldowa curves (Tech. s __ - - s Figures 3.4-2 and 3.4-3, respectiv21y). The original Sal, .ia analysiscalculatedamaximumpeakpress[ureforthemost lin';A g Thdgpj' , mass addition transient of 446 psig with the PORV set at a sg l, ; pressure of 375 psig. In this analysis, the RCS pressure due to injection of 780 gpm flow into an initially cold water solid RCS was considered. In the limiting heat addition transient, a ! maximum peak pressure of 418 psig.was calculated, j/ The NSSS vendor identified in letter PSE-93-204 dated March 15, 1993 (NSAL-93-005B) a non-conservatism in the calculation for peak pressure for the heat input and mass addition transients that affects both Salem Units 1 and 2. The concern is that the difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg pressure and the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature (P/T) limits are defined) with the reactor coolant pumps (RCP) operating was not considered in the original Salem POPS analysis. This results in encroachment on the P/T limits. To quantify the effects on Salem, specific pressure differences associated with RCP operation have been calculated for one, t w, and four RCPs operating. The results of these calculations provided values of 29, 37, and 71 psig with one, two and four RCPs operating, respectively. A correction pressure of 2.0 psig has been added to these values to account for transmitter elevation differences not previously accounted for in the calculations. As a result, procedure revisions have been implemented to limit the number of RCPs in operation to 1 while in Mode 5 (< 200*F). In addition, plant operating procedures require that the power supplies be removed from both SI pumps (675 GPM) upon entry into Mode 4 (< 350*F). Therefore, only the mass input f rom a Centrifugal Charging Pump (CCP) (560 GPM) is considered. The net results, assuming operation of a CCP and tne pressure difference from operation of one RCP, are peak pressures Falow the specified limits in the Unit 1 and 2 P/T curves. Sit.ce the completion of these evaluations, it has now been determined that the Positive Displacement Charging Pump (PDP), if already in operation, would continue to operate upon initiation of a SI signal if offsite power remains available. During this postulated event, letdown would be automatically isolated as part of the SI actuation. The additional flow from the PDP is a concern for the period of time when the RCS is < 200*F (Mode 5), the PDP is in operation, and one (1) CCP has its associated power supply available. Therefore, the combined flow of 665 GPM from the PDP (105 GPM) and the CCP (560 GPM) is now considered the most limiting mass addition transient.
r PSE&G has re-analyzed the mass addition event ysing the GOTHIC computer code assuming a bounding maximum combined pump flow rate of 675 gpm. The resulting peak pressure is 443 psig. After including the pressure differential due to the operation of one RCP and the 2 psig elevation correction, a peak pressure of 474 psig is established. This pressure exceeds the design 'oasis limit of 450 psig for Salem Unit 1. The P/T limit of 475 psig for Unit 2 continues to be met with relatively no margin. These results are summarized in the following table: UNIT POPS CALC. PEAK AP W/ 2 kCP CORRECTED T/S SETPOINT PRESS. RUNNING PEAK RCS PRESS. (PSIG) (PSIG) & ELEV. PRESSURE LIMITS - BASED ON CORRECTION (PSIG) HEATUP OR MASS INPITT (PSIG) 20*F/HR CATE W/ 6't$ CoOLDOWN GPM FLOW (PSIG) S1 375 443 31 474 450 S2 375 443 31 474 475 For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature), the original peak pressure calculated 'is 418 psig. When the pressure differential due to the operation of one RCP (29 pai) and the 2.0 psig to correct for elevation is added, the peak pressure is 449 psig. This peak pressure remains below the POPS limits of 450 and 475 psig for Salem Units 1 and 2, respectively. As. compensatory action, administrative controls ensure RHR relief valve RH3 is available and the associated RHR isolation valves i are in the open position. This action is necessary when RCS temperature is less than 200 degrees F, the PDP is in operation and a CCP has power available. Analyses for a mass addition transient up to 780 GPM, assuming either 2 PORVs, or 1 PORV and RH3 are available, have determined that sufficient relieving capacity exists to ensure that the current P/T limits will not be exceeded. This is not required if a vent path is established instead of POPS in accordance with Tech. Specs., or the PDP is not in operation. Salem Unit 1 operating procedures limit operation in Mode 5 to one RCP, require that power be removed from the SI pumps in Mode 4, and require relief valve RH3 to be available to ensure the
. current P/T limits are' met when RCS temperature is < 200*F.
Similar administrative controls for Salem Unit 2 are in place although RH3 is not needed to meet its P/T limits. However, without credit for the relieving capacity of RH3, relatively no a
a
~
operating margin between the citztant P/T limits ca.culated in accordance with 10CFR50, Appendix G and the peak pressure during low temperature overpressure transients exists for Salem Unit 2. APPARENT CAUSE OF OCCURRENCE: Based on the results of the investigation to-date, the operation of the PDP was not considered in the original POPS analysis nor was it considered by the NSSS Vendor in any other Chapter 15 accident analyses. It was believed that the PDP operation would be terminated fellowing initiation cf a safety Injection (SI) signal. Following the April 7 event at Salem Unit 1, it was determined that the PDP would continue to operate if already in service following i-itiation of a SI 11gnal with offsite power I I available. This was confirmed by a review of the current control configuration for the PDP. The cause of this omission has yet to be determined at this time. The initial POPS analysis for the mass addition transient also did not account for the differential pressure between the mid-plane of the core and the location of the pressure sensors located in the RCS '.at legs with the RCP(s) in operation. Following the April 7 event, the NSSS Vendor completed an evaluation of the continued PDP operation during SI for the complete spectrum of Chapter is accidents including the Inadvertent SI at Power transient. The evaluation also addressed the impact of continued PDP operation on affected systems and component performance. The NSSS Vendor evaluation concluded that continued operation of the PDP would not impact existing accident analyses. -However, this evaluation'did not consider the PDP operation during low temperature conditions. PRIOR SIMILAR OCCURRENCES: No ether prior similar occurrences have been identified related to this design deficiency. SAFETY SIGNIFICANCE: This event is reportable in accordance with the requirements of 10CFR50.73 (a) (2) (ii) (B) , due to the POPS not being able to meet it design basis, and provide adequate overpressure protection should a mass addition transient occur. Thin LER also satisfies any_ reporting requirements pursuant to 10CFR21. Additional margin on the Tech. Spec. curves can be gained when operating with POPS (RCS cold legs less than 312'F) by taking credit for ASME Code Case N-514. This Code Case states that the LTOP systems _shall limit the maximum pressure in the vessel to 110% of the pressure determined to a:tisfy Appendix G of ASME Section XI,-Article G-2215. Credit.ug the Code Case will allow
1 I r l the maximum allowable pressure (Tech. Spec. P/T limits) for POPS to be increased to 495 psig and 522.5 psig for Salem Units 1 and 2, respectively. The calculated peak pressure of 474 psig would be considerably below the P/T limits assuming the additional 10% margin allowed by Code Case N-514. (It is noted that the l contents of Code Case N-514 have been incorporated into Appendix G of Section XI of the ASME Code in the 1993 Addenda to Section XI.) The ASME Working Group on Operating Plant Criteria (WGOPC) developed code guidelines to define POPS limits that will avoid certain unnecessary operational restrictions, provide adequate margins against failures, and reduce the potential for unnecessary activation of pressure relieving devices used for POPS (PORVs). The philosophy used by the WGOPC for developing these guidelines was to ensure that the POPS limits are still below the P/T limits during normal plant operation, but allows the pressure that may occur with activation of the POPS to exceed the P/T limits by a maximum of 10% provided acceptable margins are mainta3ned during these events. This philosophy was to protect the reactor vessel from low temperature overpressure transients and still maintain the P/T limits in the Technical Specifications applicable for normal heatup and coo?.down in accordance with Appendices G to 10CFR50 and Section XI of the ASME Code. The WGOPC applied deterministic and probabilistic analysis techniques for several different flaw locations and heat-up and cooldown rates to establish the conditions deaineated by the Code Case. For consideration, there are several conservatisms inherent in the development of the ASME Section XI, Appendix G P/T curve calculations that include:
- 1) The safety facter of 2 on the principal membrane (pressure) stresses.
- 2) A margin factor applied to RT in accordance with the requirementsofRegulatoryGu!be1.99, Revision 2 (eg., 2-sigma margins are applied in determining the adjusted reference temperature).
- 3) The disregarding of increased mechanical properties of the reactor vessel which accompany material embrittlement (elevated yield strength and flow stress).
- 4) M1 assumed flaw in the wall of the reactor vessel that has a depth equal to 1/4 the thickness of the vessel wall and a length equal to 1-1/2 times the vessel wall thickness.
i i
7
- 5) The reference stress intensity curves defined in Section XI, Appendix G are used to bound the dynamic crack initiation and crack arrest toughness.
In summary, Code Cade N-514 allows exceedance of the P/T limits calculated in accordance with 10CFR50, Appendix G by 10%. The application of Code Case N-514 provides sufficient margin such that the inadvertent SI actuation that results in a. mass input from both a CCP and the PDP will not result in any P/T limits being exceeded assuming the availability of only one POPS. Code Case N-514 also allows the operation of up to two RCPs when RCS temperature is < 200*F, and removes the Salem Unit I requirement for RH3. Therefore, it is concluded that tnis event does not have any safety significance. CORRECTIVE ACTIONS: Salem Unit 1 operating procedures limit operation in Mode 5 to one RCP, and require that power be removed from the SI pumps in Mode 4. Relief valve RH3 is required to be available to' ensure the current P/T limits are met when RCS temperature is < 200*F when the PDP is-in operation. Relier valve RH3 is not required to be available if a vent path ;a established-instead of POPS in accordance with Tech. Specs.. Similar administrative controls for Salem Unit 2 are in plac although RH3 is not needed to meet its P/T limits. , A letter to the NRC is currently under preparation requesting application of Code Case N-514 which allows exceedance of the Pressure / Temperature (P/T) limits calculated in accordance with the requirements of 10CFR50, Appendix G by 10%. This additional 10% margin will increase the P/T limits operating. margin for the POPS during Low Temperature Ove1 pressure Protection (LTOP) conditions, and allow the operating restrictions discussed above to be removed. A License Change Request is also under consideration to credf.t RH3 as part of POPS. PSE&G is also evaluating the benefits of completing a design change to terminate PDP operation following initiation of a SI signal if offsite power remains available. The NSSS Vendor has corapleted an evaluation of the continued PDP cperation during'SI for the complete spectrum of Chapter 15 accidents including the Inadvertent SI at Power transient. The evaluation also addressed the impact of continued PDP operation on af fect ed sys 2ms and component performance. The NSSS Vendor evaluation concluded that continued operation of the PDP would not impact existing accident analyses. PSE&G has requested the NSSS Vendor to investigate any generic implications associated with the continued operation of the PDP during Safety Injection.
- w
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3 FOUR WEEK LOOK AHEAD KENNETH M. O'GARA JANUARY 17, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorptr:.t:.ng Salem ATWS long term commitments. Issued for review. With BJT (06/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization.(for Salem Update 06/30/94)
- 3. LCR for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage.
- 4. Complete reportability determination IAW Part 21, EDG Injection Studs
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Salem LCR 93-05, Safety Valves (Comments due from Salem Tech 01/24/94)
- 7. Salem LCR 93-10, Degraded Grid Voltage Setpoint (03/01/94)
- 8. HFJS Letter ;n bmitted to NRC 01/31/94
- 9. Complete Charging Pump /Boration Flow Path LCR 93-19 (04/01/94) 10 RevisiontoNAP-30basedonBiehnialReview COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Issued draft SAR changa and 50.59 to BJT for NLR Approval
- 2. Revised LCR 93-10 and Issued for Review and Comment
- 3. Comments on NAP-35, Rev. 5 to JJG AREAS OF CONCERN H2 Storage Facility letter with HC Station for 2. Months MEETINGS AND TRAINING 01/20 - 01/21 - Word Derfect Training VACATION Tentatively the last week in February (2/21-2/25)
Page 1 of 1
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FOUR WEEK LOOK AHEAD KENNETH M. O'GARA JANUARY 31, 1994 IASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Issued for review. With BJT (06/30/94)
- 2. Complete SAR change, Salem /Nuc Dept, reorganization.(for Salem Update 06/30/94)
- 3. LCR for Hope Creek from GL 92-01 res Removal in Spring 1994 Outage.(q q5) ponse. Tied to Capsule
- 4. Complete reportability determination IAW Part 21, EDG Injection Studs
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Salem LCR 93-05, Safety Valves (Resolve Comments from Salem Tech by 02/04/94)
- 7. Salem LCR 93-10, Degraded Grid Voltagt Setpoint (03/01/94)
- 8. HWCS Letter submitted to NRC 02/11/94
- 9. Complete Charging Pump /Boration Flow Path LCR 93-19 (04/01/94)
- 10. Revision to NAP-30 based on Biennial Review
- 11. Issue Letter to R&A regarding processing of Vendor Part 21's under 21.21(b) and Letter to MGMT on time clock (02/18/94) t1 .i-- . n e Chc % Yer hiu.i mpg Td 6pu.
L COMPLETED TASKS FOR PREVIOUS WEEK
- 1. 10CF;R21, Work Standard Approved .byJFXT
- 2. LCR Descript'f3ns .for/ CBLA Database completed
- 3. Reviewe,d'EDG f $tud Analysid/
Sdpported Resolution of connents on/HWCS 4. AREAS OF CONCERN ___EDG Part 21, on initial review appears to be reportable Issue regarding POPS setpoint nonconservatism may be reportable
, under 50.72.
MEETINGS AND TRAINING NONE F VACATION b Need to Still Makeup 40 Hrs. due to foul weather Tentatively the last week in February (2/21-2/25) Page 1 of 1 i 3
y 4_ _x; bf d FOUREWEEK LOOK AHEAD-
-KENNETH M.' O'GARA JANUARY M , 1994-HI-TASKS TO BE PERFORMED l'. Initiate.SAR change for incorporating Salem-ATWS long tenn = commitments. Irsued for review. With BJT -(06/30/94)-
- 2. Complete SAR change, Salem /Nuc Dept. reorganization. (for Salem Update 06/30/94)
- 3. LCR for-Hope Creek from GL 92 01 response. Tied to Capsule:
Removal in-Cpring 1994 Outage.
- 4. Complete reportability-determination IAW Part 21,-EDG Injection Studs '
5.- GL 92-01 update of HC UFSAR'regarding BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Salem.LCR 93-05, Safety Valves (Comments due from Salem Tech 01/24/94)
- 7. Salem LCR 93 10, Degraded Grid Voltage Setpoint (03/01/94)
- 8. HWCS Letter submitted to NRC 01/31/94
- 9. Complete Charging Pump /B' oration Flow Path LCR 93-19<
-(04/01/94) 10,-
g Revision to NAP-30' based on Biennial Review
- 11. Issue Letter to R&A regarding processing of Vendor Part 21's P under 21.21(b) (02/13/94) -)
COMPLETED TASKS FOR PREVIOUS WEEK j Swe o Fg &^ 6
- l-1. nitiated LG 92-?9, tcration ric N.th 2.
'l
=Assesseo Fossible changes to Vend Part 21 process y ;
W "p hted LAlefing!moon L's wtA w-ups IGr OLa 00-06 and s2-uA~ m.y
' .ms , ,n 't.-
AREAS-OF CONCERN 3 f ,, 1 _f f A~~ A H2 Storage Facility letter with HC Station for 2 Months MEETINGS AND TRAINING
-NONE VACATION g M a to Maksp tim Tt::chy for r& & b a due to foul weather 7 re&owal Day ,sEL-Tentatively the last week in February (2/21-2/25) w%su s & Wg Page 1 of 1
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FOUR WEEK'LOOK AHEAD KENNETH'M. O'GARA FEBRUARY 07,-1994
= TASKS-TO BE PERFORFID
- 1. Initiate SAR-change-for incorporating Salem ATWS long_ J
=
term commitments. Issued for-_ review. With BJT-(06/30/94) 2.. Complete SAR change,-Salem /Nuc Dept. reorganization.(for
= Salem-Update 06/30/94).
t
.- 3 .
-LCR_for. Hope Creek from GL 92-01 response. Tied to: Capsule Removal-in Spring 1994-Outage.
- 4. -Complete reportability- determi nation- IAW Part - 21, -EDG; Injection Studs
-5. GL192-01 update of HC UFSAR:regarding BWROG. Equivalent Margin Analysis-(status-by-4/30/94)
- 6. Salem LCR 93-05, Safety Valves & RERR '
- 7. Salem LCR 93-10, Degraded Grid Voltage Setpoint (03/01/94)
-8. HWCS Letter submitted to NRC 02/11/94 9: Complete Charging Pump /Boration Flow Path LCR193-19 (04/01/94) j 10 ' Revision to NAP-30 based'on1 Biennial' Review-
- 11. Issue Letter to R&A regarding processing.of Vendor Part 21's
-under 21.21(b) and-Letter to-MGMT on time clock (02/18/94)
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. LCR 93-19 draft complete. '
- 2. l Processing of Vendor Pt 21 memo J ssued for signof f.
- 3. : Memo to Mgmt. re. Part 21 to DAS for comment.
c.; 4. LIssue on POPS to NME for further review. No IR written AREAS OF CONCERN EIX3 Part- 21, on initial review appears to be reportable
~
LCR 93->6 te be. revised per Salem Comment & Draft GL of [ MEETINs '1 AND TRAINING GET Requal 2/10
-VACATION J Need to Still Makeup Hrs. due to foul weather Page 1 of 1
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- . _ - ~ ~ ~ .- -... ~ - - ,.- _ - .-. _ . . . . . . . . - . -
P f- . FOUR WEEK LOOK AHEAD TENNETH M.--O'GARA FEBRUARY 14,-1994 f TASKS:TO BE PERFORMED
- 1. . Initia'te=SAR change for. incorporating Salem ATWS long d Eterm commitments. Issued for review. With BJT (06/30/94)-
2.. _ Complete SAR-change, Salem /Nuc Dept, reorganization.(for. Salem Update 06/30/94! 3.- LCR.92-07 for Hope Creek: f rom GL 92-01'_ response. Tied .to. '
' Capsule Removal in Spring 1994 Outage. ;
- 4. Complete reportability determination IAW Part 21,..EDG Injection Studs -
'5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent T Margin Analysis (status by 4/30/94)
- 6. Salem LCR 93 05, Safety Valves & RERR
- 7. : Salem LCR 93-10, Degraded Grid-Voltage.Setpoint;(03/01/94)'
8.- HWCS Letter submitted t'o NRC 02/11/94
- 9. ; Complete Charging Pump /Boration Flow Path LCR 93 19..
(04/01/94) 10.- ' Revision to NAP-30 based on Biennial' Review
- 11. . Issue Letter to R&A regarding processing of Veador Part-21's
~ ~
under- 21.21(b): and Letter to MGMT on time' clock -(02/18/94)
- COMPLETED = TASKS'FOR PREVIOUS WEEK l '. -Drafted [EDG'PT21reportsummaryforNRCnotification. '
- 2. Processing of--Vendor PT-21 memo to FXT,
; 3. Memo to Mgmt. re. Part 21iResolved-FXT comments.
AREAS OF CONCERN' HWCS Letter still with HC Station MEETINGS AND TRAINING GET Requal-FXT Staff Meeting (; VACATION u 8 Hrs, to be made up due to foul weather Page 1 of 1 is' 3 l l
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA-L FEBRUART .21; 3994
. TASKS TO BE PERFORMED 1.- Initiate SAR change for ' incorporating Salem ATWS long term commitments. Issuedifor review.'With BJT (06/30/94)
- 2. -Complete SAR change, Salem /Nuc Dept. reorganization. (f or Salem Update- 06/30/94)-
-3. LCR 92-07'for Hope Creek from GL 92-01 response.-Tiedito Capsule Removal,in Spring 1994 Outage.
- 4. Complete reportability determination IAW Part 21, EDG~ l
- Injection Studs
- 5. GL 92-01 update of HC UFSAR regarding-BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Salem LCR 93-05,-Safety Valves &'RERR
-7. Salem-LCR 93-10, Degraded Grid Voltage Setpoint -(03/01/94):
- 8. HWCS Letter-submitted;to-NRC 02/11/94
- 9. -Complete Charging Pump /Boration Flow' Path LCR 93 (04/01/94)
- 10. Revision to NAP-30 based on Biennial Review 111.- Issue Letter to R&A-regarding processing of Vendor Part 214s under 21.21(b) and' Letter to MGMT on time . clock (02/18/94)
COMPLETED TASKS FOR PREVIOUS' WEEK-
.1. 'EDG PT 21 cceplete - Not Reportable.
2.- Resolve FXT comments ~on Vendor PT 21 memo.
- 3. -Interface with S.-Morris L(NRC) to close=2 Vios. -.i JGEJG OF CONCERN l HWCS-Letter still with HC Station - K. Maza Need to discuss Pt 21 Mgmt mern.'Is it needed?
Need to discuss RERR Semi-annual Report-LCR MEETINGS AND TRAINING None VACATION Still Hrs.-to be made up due to foul weather Page 1 of 1
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t FOUR WEEK LOOK AHEAD KENNETH M. O'GARA FEBRUARY 28, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments.-Issued for review. With BJT (06/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization.(for.
Salem Update 06/30/94)
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage.
4.
- 5. GL-92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Salem LCR 93-05, Safety Valves & RERR
- 7. Salem LCR 93-10, Degraded Grid Voltage Setpoint (03/01/94) l 8. HWCS-Letter submitted to NRC 03/11/94
- 9. Complete Charging Pump /Boration Flow Path LCR 93-19 (04/01/94)
( 10. Revision to NAP-30 based on Biennial Review I
- 11. Issue Letter to R&A regarding processing of Vendor 1 Part 21's under 21.21(b).
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Part 21.21(b) memo to FXT for sign-off 2nd round of comments incorporated
- 2. PT 21 Mgmt memo cancelled per disc. w/ DAS
- 3. Address Fuel comments on Charging PUMP LCR
- 4. Revised HWCS letter.to address HC Chem comments AREAS OF CONCERN
.HWCS Letter still with HC Station - K. Maza Need'to-discuss RERR Semi-annual Report LCR MEETINGS AND TRAINING
-None VACATION Still 3 Hrs. to be made up due to foul weather r ll Page 1 of 1 !
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FOUR WEEK LOOK' AHEAD' KENNETH M. O'GARA MARCH 7,_1994'
-TASKS-TO BE-PERFORMED
- 1. InitiateLSAR change for incorporating Salem ATWS long iterm commitments. Issued forireview. With BJT (06/30/s.
- 2. Complete SAR change, Salem /Nuc Dept. reorganization.(for-Salem-Update 06/30/94)
- 3. LCR 92-07 for-Hope Creek _from GL.92-01 response. Tied to Capsule Removal-in Spring 1994 Outage.
- 4. Potential LER regarding POPS setpoint analysis 5,
GL 92-01 update.of HC UFSAR regarding BWROG Equ$ valent Margin Analysis (status by 4/30/94)
- 6. Part 21 60 day eval of Inadvertent ~SI Operator Response
- 7. Salem LCR 93-10, Degraded Grid Voltage Setpoint (03/18/94)-
- 8. HWCS Letter submit to NRC 03/11/94=
- 9. Complete Charging Pump /Boration Flow Path LCR 93-19 1
-(04/01/94) 10 . . Revision to NAP-30 based on Biennial Review
- 11. Part 21 60 day eval. of incorrect fuel pellet (5/3/94). >
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Part 21.21(b)-' memo issued
- 2. Drafted Pt21 Memo re. inadvertent SI Operator Response-
- 3. Drafted-Pt21 Memo're. incorrect fuel pellet
- 4. SEM request on Sig. Event PI AREAS OF CONCERN HWCS-Letter still-with HC Station - K. Maza' Potential LER on POPS setpoint analysis MEETINGS AND TRAINING None
^
-VACATION
-Still 1 Hr.-to be made up due to foul weather l
Page 1 of 1 f-
i FOUR WEEK LOOK AHEAD KENNETH M. O'GARA MARCH 14, 1994 TASKS TO BE PERFORMEQ
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Issued for review. With BJT (06/30/94)
- 2. Complete SAR change, Salem /Nuc_ Dept, reorganization.(for Salem Update 06/30/94)
- 3. LCR 92-07 for Hope Creek from GL,92-01 responses Tied to Capsule Removal-in Spring 1994 Outage. '
- 4. Potential LER regarding POPS setpoint analysis
- 5. .GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Part 21 60 day eval of Inadvertent SI Oparator Response
- 7. Salem LCR 93-10, Degraded Grid Voltage Setpoint (03/18/94)
- 8. HWCS Letter submit'to NRC 03/11/94
- 9. Complete' Charging Pump /Boration Flow Path LCR'93-19 (04/01/94)
- 10. Revision to NAP-30 based on Biennial Review
- 11. Part 21 60 day eval, of incorrect fuel pellet (5/3/94).
COMPLETED TASKS FOR PREVIOUS WERE
- 1. Issued LCR 93-19 for comment
- 2. Issued Pt21 Memo re. incorrect fuel pellet
- 3. Issued Memo on NRC Review Team Report (50.7)
- 4. HWCS letter for DAS sign-off AREAS OF CONCERN Potential LER on POPS setpoint analysis Part 21 Memo on inadvertent SI event needs to be issued l_ MEETINGS AND TRAINING SORC 03/16/94 -
LCR 93-10 VACATION NONE 4 Page 1 of 1 .
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i FOUR WEEK-LOOK AHEAD KENNETH M. O'GARA MARCH 21, 1994
~ TASKS TO'BE PERFORMED; _
- 1. Initiate.SAR change-for incorporating: Salem ATWS longL term' commitments. Issued for review. With BJT--(06/30/94)
- 2. LComplete'SAR_ change,-* Salem /Nuc? Dept. reorganization'. (for.-
Salem Update l06/30/94)
- 3. <-LCR 92-07-;for._ Hope Creek from GL 92-01-response. Tied-to Capsule Removal'~in Spring 1994 Outage.
I
-4i Potential LER-regarding POPS lsetpoint analysis
- 5. GL 92-01 update of HC UFSARfregarding.BWROG' Equivalent-Margin Analysisi(status by 4/30/94)
- 6. Part-21 _60 day, eval of Inadvertent SI Operator Response (5/8/94)
.7. Salem LCR 93-10,--Degraded Grid Voltage Setpoint:f(03/25/94)'
8.- -HWCS Letter: submit lto NRC 03/31/94~ D. Complete Charging < Pump /Boration Flow Path LCR 93-19' _(04/01/94) 110. Revision to NAP-30' based on Biennial Review
- 11. Part 21 day eval, of . incorrect fuel _ pellet :(5/3/94) .
COMPLETED-TASKS FOR PREVIOUS' WEEK-
.1. - Resolved DAS-comments on HWCS Letter-
- 2. Special Task for:FXT re. Licensing Database Potential LER^on POPS'setpoint analysis. DEF to be written,
'3.-
- 4. -Issued inadvertentLSI PT21 memo AREAS OF CONCERN NONE MEETINGS AND TRAINING
-SORC 03/23/94 - LCRl93-10 VACATION POSSIBLY 6/3 -
6/13
'Page 1 of 1 ,
a/L
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@~
g .y :FOUR WEEK:LOOK AHEAD KENNETH M.'O'GARA -
, MARCH 28, 1994.-
-TASKS TO BE'PERFORMEQ--
L 1. -: Initiate SAR; change.forJincorporating= Salem ATWS long
- term commitments. Issued for. review.1With BJT.l'(06/30/94)
- 2. Complete SAR change, Salem /Nuc Dept.--reorganization.(for
--Salem 1 Update 06/30/94):-
i-l 3 '. lLCR _ 92 - 07 : f orJ Hopel Creelt from GL-92-01iresponse.. Tied to Capsule' Removal in Spring'1994. Outage.
-4. -Potential LER:regarding. POPS setpoint> analysis'
- 5. -GL 92_01 update _of HC UFSAR regarding'BWROG' Equivalent.-
Margi.n: Analysis- (status by 4/30/94)- 6 ~. P. art 21~60 day _ eval ofLInadvertent SI Operator-Response-
.(5/8/94)
- 7. Salem LCR 93-10,' Degraded Grid Voltage:Setpoint-(03/25/94) . .
- 8. HWCS Letter submit 1to~NRC 03/31/94
- 19. LComplete Charging Pump /Borati~on: Flow Path LCR $3-19 (04/01/94)
- 10. Revision'to NAP-301 based'on Biennial Reviews
- '11.. ,Part 21 60 day eval, of incorrect-fuel.l pellet.(5/3/94).
COMPLETED TASKS-FOR' PREVIOUS-WEEK ^ SORC'd 2 LCRs 1.
.2- .
Issued. letter on-50.7 review team report
- 3. Mtg on POPS setpoint analysis. DEF_ reviewed
'4. -Resolved' Charging Pump-LCR comments
, AREAS OF CONCERN ~ ' LCharging Pump LCR still awaiting Tech'& NSR' comments SAR Changes:for Salem Update Update Frozen?
@ ETINGS AND TRAINING.
IM en fd 2I 9toda
-VACATION-POSSIBLY 6/3 - 6/13 Page 1 of 1
, , t.
s; - FOUR WEEK LOOK AHEAD KENNETH M. O'GARA
. APRIL 11, 199^
TASKS TO BE PERFORMED
- 1. - Initiate SAR change for incorporating Salem ATWS long term commitments. Issued for review. With BJT-(06/30/94)-
-2. Complete SAR change, Salem /Nuc Dept. reorganization.(for Salem Update 06/30/94)
- 3. LCR 92-07.for Hope Creek from GL 92-01 response. Tied to i -Capsule Removal in Spring 1994 Outage.
- 4. Potential LER regarding POPS setpoint analysis
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (status by 4/30/94)
- 6. Part 21 60 day eval of Inadvertent SI Operator Response (5/8/94)-
- 7. Complete Charging Prq /Boration-F1 i Path LCR 93-19 (04/01/94)
- 8. Revision to NAP-30 based-on Biennial Review
- 9. Part 21 60 day eval of incorrect fuel' pellet-(5/3/94).
COMPLETED TASKS FOR~ PREVIOUS WEEK ~
- 1. HWCS Letter submitted to NRC 2; Resolved LCR 93-19 comments
- 3. 21.21(b) procedure rev. request i
- 4. Followup on SORC comment for BF22s
- 5. Part 21'on Fuel Pellet for signoff
- 6. Industry contacts for PORV Violation i
AREAS OF CONCERN DEF on POPS Setpoint Nonconservatism MEETINGS AND TRAINING NONE VACATION 6/3 - 6/13 Page 1 of 1 r L
u
- FOUR WEEK.LOOK AHEAD._
'KENNETH M.~O'GARA APRIL 18,11994-
- IAJ,MS -TO BE PERFORMED s l'. ;Initdiate SAR change for incorporating ' Salem ATWS long:
Eterm commitments. Issued for. review. With BJT (06/30/94)- l 2 '. - CompleteT SAR* change,- _ Salem /Nue Dept .L reorganization.-(f or Salem-Update 06/30/94)=
- 13. LCRp92-07(for Hope Creek.from GL 92i O1-response'. Tiedito-
-Capsule ~ Removal in-Spring-1994 Outage, f4. Potential LER:regarding-POPS'setpoint_ analysis L 5 ~. GLI 92-01_ update of HC UFSAR-regarding.BWROGLEquivalent l Margin Analysis-(status by-4/30/94)
- 6. Part:21;60 day eval oft Inadvertent SI Operator Response (5/8/94).
1
- 7. Complete = Charging Pump /Boration Flow Path LCR 93 .(04/22/94)L
- 8. -Revision to! NAP-30 based-on Biennial Review' COMPLETED TASKS FOR' PREVIOUS WEEK.
- 1. LCRi93'19< ready for SORC
~2. -Part 211on Fuel Pellet-complete
=3. 2 days training-on-Wordperfect
- 14. Begin draft response to Inad. SI Part -AREAS OF CONCERN-
;DEF on[ POPS Setpoint Nonconservatism Need meeting on 21.21(b) procedure rev -request' .
MEETINGS AND TRAINING SORC 4/20/94-- VACATION 6/3-- 6/13 i s
. Page 1 of 1 ,
,ir j
y *" FOUR WEEKJLOOK AHEAD-KENNETH M..O'GARA Mayf02,J1994
- LTASXS:TO~BE PERFORMED 1
!1. Initiate SARcchange:for incorporating-Salem ATWS long' term commitments.LIssued-for' review. With1BJT--(06/30/94)
~
2.: ' Complete _ SAR change, ' Salem /Nuc _ Dept; -reorganization. (for Salem! Update-06/30/94).
'3,
.LCR.-92-07:for1 Hope Creek from GL 92-01Lresponse. Tied to.
'apsule? Removal 11n Spring 1994 Outage..
- 4 '. Potential;LER,regarding POPS:setpoint analysis
- 15. lGL;92-01 update of HC~UFSAR regarding BWROG~ Equivalent
' Margin Analysis (statusLby 4/30/94)- j
, 6. Part 211 60 day eval offInadvertent SI Operator Respense-(5/8/94):
. --7 . CompleteICharging Pump /Boration-Flow Path LCR 93-19 (05/14/94) e:
8.- Revision to NAP-30ibased on Biennial Review-
- 9. 1 Response to ASTA-Engineering-Allegation (05/25/94)
'10. LGL 92-01 Let'er to-NRCL(05/13/94) '
- 11. Part'21194-03, High: Steam LineLFlow Signal"Summators (06/28/94)
- 12. R21".21~(b) procedure changes _and resolution
-COMPLETED' TASKS-FOR PREVIOUS WEEK
,1l. Final Draft:PT"21'94-01. ready for:signout
. 2. Part;21~94-03 letter on SignalfSummators.to DAS
- 13. Draft-GL 92-01 Response
=4.. PORVsLicensing/ Design-basis whitepaper AREASiOF CONCERN DEF!an,POPSLSetpoint Nonconservatism. Awa' ting RH3 analysis
' Need meeting.cn1 21.21(b) procedure rev.'raquest
' MEETINGS AND TRAINING
- Meeting 04/28/94 on Inadvertent SI Part 21 VACA7 Zd 6/3 '-_6/13 Page 1 of 1 i -
.i..i - a
r)
-FOUR WEEK LOOK AHEAD =
LKENNETH M. O'GARA May-_09, 1994-
, , TASKS-TO BE' PERFORMED < -
- 1. l Initiate SAR. change for'ine'orporating Salem ATWS longr term commitments. Issued for-review.-With_BJT (06/30/94) 27 Complete SAR-change,.: Salem /Nuc' Dept. reorganization. (f or SalemLUpdate 06/30/94)
~3. -LCR--92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994_ Outage,
'.;vu nti' sqF'^
-4. -
rR_regarding. POPS setpoint-analysis. i
. 5. NGL 92-01. update-of HC UFSAR regarding BWROG Equivalent-
. Margin _ Analysis -(status by i /E/M) l IlhV -
-6. Part-21 94-04 60 day. eval,_ FORV Plug / Stem Material (7/a/94)
- 7. -Complete Charging Pump /Boration Flow Path LCR 93-19--
-(05/14/94)
- 8. Revision to: NAP-30 based on Bi'nnial e Review 9.- Response'to ASTA1 Engineering Allegation (05/25/94)
, -10... GL.92-01' Letter to NRC (05/13/94) li. Part.21 94-03,_ High Steam-Line-Flow Signal Summators (06/28/94)
- 12. :21.21(b)' procedure. changes..and. resolution- R
' COMPLETED TASKS'FOR PREVIOUS WEEK l'.
PT 21 94-01 closed -LNot Reportable: J
-2. 'Part 21 94-04 letter on-PORV Plug Stem material to DAS
. -3. GL'92-01 Response to~FXT-
'A'REAS OF CONCERN t
'DEF on PODS Setpoint Nonconservatism. Awaiting RH3 analysis
. Need meeting on. 21.21 (b) pt:cedure rev. request ASTA Engineering Input From R. Swanson-late MEETINGS AND TRAINING NONE '
VACATION 6/3'- 6/13 . Page 1 of 1 t e,
FOUR WEEK LOOK AHEAD-KENNETH M. O GARA May 16, 1994 TASKS TO BE PERFORMEQ
- 1. Initiate SAR' change for incorporating Salem-ATHS long term commitments. Issued for review, With BJT (06/30/94)
- 2. Complete SAR_ change, Salem /Nuc Dept. reorganization;(for Salem Update 06/30/94)-
- 3. LCR-92 07 for Hope Creek from.GL 92-01 response. Tied to Capsule _ Removal in Spring 1994 Outage.
4.- DEF regarding POPS setpoint analysis. Need LCR.
- 5. GL 92-01 updateLof HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95)
- 6. Part 21 94-04 60 day eval, PORV Plug / Stem Materiai (7/8/94)
- 7. Complete Charging Pump /Boration Flow Path LCR' 93-19 (05/31/94)
Revision to NAP-30-based on Biennial Review '
- 9. Response to ASTA Engineering Allegat' ion (05/25/94)
- 10. Part 21-94-03, High Steam Line Flow Signal Summators (06/28/94)
;11. 21.21(b) procedure changes and resolution.
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Part 21 94-04 letter on PORV Plug Stem. material issued
- 2. GL'92-01 issued to NRC
- 3. Reviewed QA Input for ASTA Letter
. AREAS OF CONCERN CCP LCR - Need to discuss w/ R. Villar Need meetingoon 21.21(b) procedure rev, request.
Possibly 2 Additional Part 21's to be issued this week MEETINGS AND TRAINING
-NONE VACATION 6/3 - 6/13 (NOTE: Could Be Cancelled)
Page 1 of 1
, ' A : ':
__o
- , I FOUR WEEK LOOK AHEAD KENNETH M. O'GARA May 31, 1994-
-TASKS TO-BE PERFORMED
- 1. - Initiate SAR change:for incorporating Salem ATWS long term commitments. Issued for review. With DJT (06/30/94)
- 2. Complete-SAR change,; Salem /Nuc Dept. reorganization.(for Salem Update 06/30/94)
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to
~
-Capsule Removal in Spring 1994 Outage.
- 4. DEF-regarding POPS setpoint analysis. Need LCR.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95)
- 6. Part 21 94-04 60 day eval, PORV Plug / Stem Material (?/8/94) ,
- 7. Complete Charging Pump /Boration Flow Path LCD 93-19 (05/31/94)
- 8. Revision to NAP-30 based son ' Bier.nial Review
- 9. Part 21 94-05, Moore Booster-Relf.ys (07/01/94)
- 10. Part 21.94-03, High Steam Line-Flow Signal Summators (06/28/94) 11, 21.21(b) procedure changes _and resolution
- 12. Part 21 94-06, BDI Damper Bearings-(07/18/94)
- 13. GL 92-01 (Salem) confirmation letter (06/20/94)
~ . COMPLETED-TASKS FOR PREVIOUS WEEK
. 1. Part 21 94 letter on Moore Booster Relays ist.ued
-2. HC GL 92-01 supplemental letter (Fluence) for signature
- 3. ASTA Letter issued to NRC
- 4. Part 21-94-06 letter on DBIDs issued
- 5. LCT 93-10 letter transmitting cales. to NRC for signoff
- 6. Info on Thrust Coeeficients for pipe whip forces AREAS OF CONCERN CCP LCR - Need to discuss w/ P. Ott Need meeting on 21.21(b) procedure rev. request MEETINGS AND TRAINING Meeting w/ NME on POPS setpoint VACATION 6/3 -
6/13 See Separate Status Memo L
FOUR WEEK LOOK AMERD KENNETH M. O'GARA
. Juna 27, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Issued for review. Witin" " ' '
~ ~' 41
.fh 2.
Complete SAR change, Salem /Nuc Dept. reorganization.(w/ RAR)
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage.
- 4. DEF regarding POPS setpoint analysis. Need LCR and Code Case Letter.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95) 6.
Part 21 94-04 60 day eval, PORV Plug / Stem Material (7/8/s4)
- 7. Complete Charging Pump /Boration Flow Path LCR 93-19 (07/11/94)
- 8. Revisica to NAP-30 based on Biennial Review ..
- 9. Part 21 94-05, Moore Boostar Relays (07/01/94)
- 10. Part 21 94-03, High Steam Line Flow Signal Summators (06/28/94) 4 11, 21. 21 (b) procedure changes and resolution
>hgk 1
Part 21 94-06, BDI Damper Bearings (07/18/94) 12.
\
7-,., ,y N 13. GL 92-01 (Salem) confirmation letter (07/20/94) g' _ -- COMPLETED TASKS FOR PREVIOUS WEEK ,.[
~
. 1. Part 21 94-05, Moore Booscer Relays to FXT ~
w ,, ; 3 '
- 2. Sal GL 92-01 supplemental letter issued .
- 3. Part 21 94-04, PORV plug / stem, memo drafted / ,b "
, g/s..
- 4. Part 21 94-03, Summators to FXT e Database search for PXT on DCPs 5.
a
- 6. Response to QA Surveillance Report 94-145 ', '
/ '
/
g - AREAS OF CONCERN '
/ ,
CCP LCR - Need to discuss w/ P. Ott _,7~. i MEETINGS AND TRAINING NONE ,.. f VACATION '
/- e.
.g ' /
? I None 'p\ /
., l Page 1 ou .
y - .-,, o
_b FOUR WEEK LOOK RHEAD KENNETH M. O'GARA July 05, 1994 TASKS TO BE PERFORMED 1. Initiate SAR change for incorporating Salem ATWS long tenm commitments. Issued for review. With BJT'(06/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. . reorganization.(w/ RAR)
- 3. LCR 92-07 for Hope-Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. '
- 4. DEF regarding-POPS setpoint analysis. Need LCR and Code-Case Letter.:
5. GL.92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95)
- 6. Part 21 94-04 60 day eval, PORV Plug / Stem Material (7/8/94)
- 7. Complete Charging Pump /Boration Flow Path LCR 93-19 (07/11/94)
-8.
Revision to NAP-30 based on Biennial Review (01/31/95)
- 9. Submit RAI response to'NRC on Lic. Appl. Falsif. (07/22/94)
-10. 'Part 21 94-03,-High Steam Line Flow Signal Summators report to NRC (07/29/94)
- 11. 21.21(b) procedure changes and resolution
- 12. Part 21 94 06, 60 fr.y eval BDI Damper Bearings (07/18/94) 13.
GL 92- 01. (Salem) confirmation letter (07/20/94) COMPLETED TASKS FOR PREVIOUS WEEK
. 1. Part 21 94-05, Moore Booster Relays closed
- 2. Began Draft of F042 letter to NRC
-3. Pav. el 94-04, PORV plug / stem, to FXT
- 4. Part 21 94-03, reported to NRC
- 5. Database search for FXT on DCPs
;_REAS OF CONCERN CCP LCR - Need to discuss w/ P. Ott MEETINGS AND TRAINING NONE VACATION None
( Page 1 of 1 f
/
- FOUR WEEN LOOK~ AHEAD KENNETH M. O'GARA July 11, 1994 TASKS TO BE PERFORMED
- 1. - Initiate SAR change for' incorporating Salem ATWS long 4
-term commitments. Issued for review.-With RAR-(12/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization.(w/ RAR) 3 .' LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 2994 Outage.
- 4. DEF regarding POPS setpoint analysis. Need LCR and Code Case Letter.
- 5. GL 92-01 update of HC UFSAR regarding 3WROG Equivalent Margin-Analysis (1/31/95) .
6.- Part 21 94-04 60 day eval, PORV Plug / Stem Material
'7. Complete Charging Pump /Boration Flow Path LCR 93-19 (08/30/94)
- 8. Revision to NAP-30 based on Biennial Review (01/31/95)
- 9. Submit RAI response to NRC on Lic. Appl. Falsif . - (07/22/94)-
- 10. Part 21 94-03, High Steam Line Flow Signal Sunnators report to NRC (07/29/94)
- 11. 21.21(b) procedure changes and resolution
- 12. Part 21 94-06, 60 day eval BDI Damper Bearings (07/18/94)
- 13. GL 92 (Salem) confirmation letter (07/20/94)'
COMPLETED TASKS FOR PREVIOUS WEEK
, 1. ' Drafted Letter on GL 92-01
- 2. Part 21 94-04, PORV plug / stem to SLB (5 Work days)
- 3. - Part 21 94-03, Summators-30 day report drafted
- 4. Database' report for FXT on DCPs completed AREAS OF CONCERN CCP LCR -
Need to discuss w/ P. Ott-Input for GL 92-01 reponse due 07/13/94 BDID input due 07/11/94 MEETINGS AND TRAINING NONE VACATION None Page 1 of 1
-FOUR WEEK LOOK AHEAD KENNETH M. O'GARA
-July 18, 1994 TASKS TO BE PERFORMED-
- 1. Initiate SAR change.for incorporating Salem ATWS-long term commitments. Issued for review. With RAR (12/30/94)
-2. Complete SAR change, Salem /Nuc Dept. reorganization.(w/ RAR)
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal'in Spring 1994 Outage.
- 4. DEF.regarding POPS setpoint analysis. Need LCR and Code Case Letter.
- 5. GL.92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95)
- 6. Part 21 94-04 30 day _ report, PORV Plug / Stem Material (8/12/94)
- 7. Complete Charging Pump /Borntion Flow Path LCR 93-19 (08/30/94) 8.
Revision to NAP-30 based on Biennial Review (01/31/95)
- 9. Submit RAI response to NRC on Lic. Appl. Falsif. ( 07/22/94) ;
- 10. Part-21 94-03, High Steam Line Flow Signal Summators report to NRC (07/29/94)
- 11. 21. 21 (b)' procedure changes and resolution 12.
GL 92-01 (Salem) confirmation letter (07/20/94)
- 13. Submit letter to NRC, Degraded Grid Voltage RAI (8/30/94)
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Letter on GL 92-01 in signoff
- 2. Part 21 94-04, PORV plug / stem reported to NRC
- 3. Part 21 94-06, BDIDs, closed - Not Reportable
- 4. SRO Falsification letter in signoff
- 5. Part 21 94-04 30 day report in signoff
- 6. Telecon with NRC, Degraded Grid Voltage AREAS OF CONCERN CCP LCR - Need to discuss w/ P. Ott. Will not return call MEETINGS AND TRAINING NONE VACATION ,
None Page 1 of 1 .
. , , , . . _ . .. . ~ . - - - .. - - . .-~ , .. ~ ~ - . . _ . . . - - - ~
g , FOURLWEEK?LOOK AHEAD i KENNETH M.-O?GARA-iJULY 25f 1994
- TASKS TO BE PERFORMED ,
L1. - Initiate-SAR change for. incorporating. Salem ATWS-long-tierm - commitments. - Draf t. SAR change;Issuedi for review : Need to: complete 150.59/ Safety Evaluation;in-support offchange_ (12/30/94) -
'O. :
Complete: SAR change, Salem /Nuc. Dept. reorganizationifor
; Salem andiHope : Creek _.: _ Draft change ands 50.59-Safety.
t E , evalution:with RAR forfreview and comment by_ nuclear _
- department. .
- 3. - LCR'92-07 .for: Hope' Creek from'GL 92-01'responre. Tied'to
~
Capsule. Removal injSpring.1994: Outage. Current schedule ~:- ie '
- 4th Q 1994 :for preparation _.of LCR. +
l
'4. DEF regarcing POPS setpoint analysis. Need LCR_to creditnRH3 '
Relief Valve. . To address'QA surveillance finding, minor _TS E
- change for the max. P-T curve heatup-rate wil1 also be-addressed. Also', c9ed to prepare, letter'to.NRC. requesting-C approval of Code Case-to extend P-T curves lot during LTOP-- '
conditions. l 5.. GL:92.01. update of HC.UFSAR;regarding-BWROGJEquivalent. - Margin' ' Analysis . -(1/ 31/95 ) ._ .. Submit justification for useiof ; EOL USE values to NRC based on; CEOG :ef forts. (12/31/95): , s 6. - Complete Charging Pump /Boration Flow Path LCR 93-19l.~
-.(08/3 0/94) . Resolve; operations comments'.and'SORC.
- 7. ' Revision to NAP-30' based'on' Biennial Review (01/31/95)
~
- 8. Part 21"94-03,-High1 Steam Line. Flow Signal-Summators l=
30 day report to NRC '(07/29/94). .
- 9. 21.'21_(b)-' procedure changes and.- resoluti'on.
t 110. Submit: letter to NRC,-Degraded Grid Voltage RAI per c discussion with the NRClon July 13. Need input from NEE. (8/30/94) Submit letter to NRC-_on HPCI pump-torus isolation valve.F042
~
- 11. +
- per'50.59-to address automatic--opening following reset of isolation signal. Meeting:to discuss with I&C is csheduled g
p -' - _q for 7/26/94.18/30/94) - 1 L Page 1 of 2 , G
,y-, --
FOUR WEEK LOOK AHEAD KENNETH M...O?GARA JULY 25, 1994 COMPLETED TASKS FOR PREVIOUS-WEEK
.1. -Supplemental Letter on GL 92-01 (Salem) submitted to the NRC on 7/20/94.
Part 21 94-04, PORV plug / stem 30 day-letter-submitted to the 2 .- - NRC1on-7/25/94. No further action-is required-
- 3. Provided background Info to NRC on Signal Summator Part 21
- 4. SRO application-Falsification letter-submitted to.NRC on-7/25/94.
- 5. -Part 21 94-03, Signal Summator 30 day report with SLB AREAS OF CONCERN.
CCP LCR Need to discuss w/ P. Ott. Will not return call-ME.ETINGS AND TRAINING PDP Operation during Post-LOCA recire. 7/'.1/94 VACATION None Page 2 of 2 i e h
m, .
*/
-FOUR WEEK:LOOK AHEAD KENNETH M. O'GARA.
. AUGUST ';1, :1994'
' TASKS TO'BE PERFORMED- '
- 1. - -Initiate SAR change:for incorporating = Salem ATWS long term commitments. Draft1 SAR change Issuedifor review. Need to complete 50.591 Safety Evaluation in support of change
-(12/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization for_-
Salem and-Hope _ Creek. Draft change'and 50.59 Salety evalution-with RAR for review and comment.by nuclear department.
- 3. LCR 92-07 for Hope Creek from=GL 92-01-response. Tied;to Capsule Removal'in Spring:: 199410utage._l Current schedule!is 4th Q 1994 for preparation of LCR.
~4. DEF regarding POPS setpoint: analysis. Need LCR to credit RH3-Relief Valve.- To address QA surveillance finding,cminor TS
,3- _ change for the max. P-T curveLheatup' rate will also be addressed. Also, need-to prepare letter to NRC requesting approval of Code Case to extend P-Tl curves 10% during LTOP conditions. 5.. -GL 92-01 update _of HC UFSAR regarding BWROG Equivalent = Margin-Analysis (1/31/95).~ Submitfjustification for use of-EOL USE values to=NRC based-on CEOG_ efforts.(12/31/95). I. f6.- Complete Charging Pump /Boration Flow Path-LCR 93-19 (08/30/94). Resolve operations comments and SORC. 7. Revision to NAP-30fbased on Biennial Review (01/31/95)
- 8. Submit let'ter to NRC addressing NJDEP comments on Hydrogen Water Chemistry-letter. Need to address administrative-controls-forLflashfflooding and how:the trailers are- l
. presently secured to thenground.
- 9. 12'1. 21 (b) procedure changes and resolution
- 10. Submit letter to NRC, Degraded Grid Voltage RAI per~
-discussion with.the NRC on-July 13. Need input.from NEE.
(8/30/94) , 11.1 Submit =-letter.to NRC on HPCI pump torus isolation valve;F042 _per.50.59 to-address automatic opening following reset of isolation. signal. Met _with I&C. Draft letter provided,to I&C. I&C will_ provide input for F031 valve and why it is different than F042 (8/30/94)
- 12. Member of Fatigue Mon toring System Project Team. Following installation of the=new system, a LCR will be required to take credit for the more accurate fatigue usage factor.
Fage 1 of 2 1
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA AUGUST 01, 1994 TASKS TO BE PERFORMED (Cont'd)
- 13. Address SERT 94-04 Open Item to resolve inadvertent SI at Power issue and PDP' operation following SI with No loss-of_offsite power. Need to SORC UFSAR change by.
end of-september. Draft SE by H is currently under review. COMPLETED TASKS FOR PREVIOUS WEEE
- 1. Completed review of Fatigue Monitoring System PSP
- 2. Completed review of W draft SE for Inadvertent SI at Power and PDP Operation to address SERT 94-04 Open Item
- 3. Draft response to_NJDEP comments on HWCS prepared and issued for review and comment
- 4. Issued draft F042 letter to I&C for input and review
- 5. Part 21 94-03, Signal Summator 30 day report issued to NRC AREAS OF CONCERN CCP LCR - Need to discuss w/ P. Ott. Will not return call Revised letter to NRC on Part 21 - Comsip Microswitches may be needed MEETINGS AND TRAINING Meeting with I&C to discuss F042 submittal (7/26).
Lunch Time Technical Topic Mtg. on Rx Vessel Structural Integrity (7/28) VACATION Plann!.lg to Take 7/15/94 as Vacacion Page 2 of 2 / ( I
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA AUGUST 08, 1994 TASKS TO BE PERFORMED
- 1. . Initiate SAR change for incorporating Salem ATWS long ;
term commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
'2. Complete 9AR change, Salem /Nuc Dept, reorganization for Salem and Hope Creek. Draft change and 50.59 Safety evalution with RAR for review and comment by nuclear department.
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3
! Relief Valve. To address QA surveillance " nding, minor TS change for the max. P-T curve heatup rate w.ll also be addressed. Also, need to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions. f
- 5. GL 92-01 update of HC UFSAR regarding BWROG. Equivalent Margin Analysis (1/31/95). Submit justification f or use of EOL USE values to NRC based on CEOG efforts.(12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19 (08/30/94). Resolve. operations comments and SORC.
7. Revision to NAP-30 based on Biennial' Review (01/31/95)
- 8. Submit letter to NRC addressing NJDEP comments on Hydrogen Water Chemistry letter. Need to address administrative controls for flash ficoding and how the trailers are presently secured to the ground.
- 9. 21. 21 (b) procedure changes and resolution
- 10. Submit letter to NRC, Degraded Grid Voltage RAI per discussion with the NRC on July 13. Need input from NEE.
(8/30/94)
- 11. Submit letter to NRC on HPCI pumo torus isolation valve F042 per 50.59 to address automatic opening following reset of isolation signal. Met with I&C. Draft letter provided to I&C. I&C will provide input for F031 valve and why it is different than F042 (8/30/94)
- 12. ' Member of Fatigue Monitoring System Project Team.
Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor. Page 1 of 2 j j j
FOUR, WEEK LOOK AHEAD KENNETH M. O'GARA AUGUST 08, 1994 TASKS TO BE PFRFORMED (Cont'd)
- 13. Address SERT 94-04 Open Item to resolve inadvertent SI et Power issue and PDP operation following SI with No
-loss of offsite power. Need to SORC UFSAR change-by end of september. Drat SE by 1! is currently .under -
review. COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Resolved comments on HWCS letter.-Issued for signature.
- 2. Prepared memo to Part 21 94-01 File re Inadvertent :iI at Power event per FXT request
- 3. Reviewed Enforcement File for commitments related to j DCP process.
- 4. F042 letter to NRC-issued for signature.
- 5. LCR 93-10 letter drafted. Need input from NEE for 2 questions.
- 6. Provided response to R&A on spurious closure of CCP
- 7. recirc. line isolation valves for Salem 1 & 2.
Drafted 50.7 memo for Nuc. Depts. to roll down. AREAS OF CONCERN CCP LCR - R. Villar to discuss w/ P. Ott. Revised be needed letter to NRC on Part 21 - Comsip Microswitches may MEETINGS AND TRAINING NONE VACATIO!{ Planning to Take 7/15/94 as Vacation t b Page 2 of 2 1
/
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA AUGUST PO 1994 If IASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for_ review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization for Salem and Hope Creek. Draft change and 50.59 Safety evalution %ith RAR for revi9w and comment by nuclear department. ,
- 3. LCR 92 07 for Hope Creet from GL 92-01 response. Tied to Capsule Remo.'al in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 nelief Valve. To address QA surveillance finding, minor TS change for the max. P T curve heatup rate will also be addressed. Also, need to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions.
- 5. GL 92 01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values to NRC based on CEOG efforts. (12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19 (08/30/94). Resclve operations comments and SORC.
.ser-es L _
7.
~
Revision to NAP 30 based on Biennial Review (01/31/95)
- 8. Submit letter to NRC addressing NJDEP commer's on Hydrogen Water Chemistry letter. Need to address administrative controls for flash flooding and how the trailers are presentlysecuradtotheground.($hu)
- 9. 21. 21 (b) procedure changes and resolution
- 10. Submit letter to NRC, Degraded Grid Voltage RAI per discussion with the NRC on July 13 Need 1nput from NEE.
(8/30/94)
- 11. Submit letter to NRC on HPCI pump torus isolation valve F042 per 50.59 to address automatic opening following reset of isolation signal. Met with I&C. Draft letter provided to I&C. I&C will provide input for F031 valve and'why it is different than F042 (8/30/94)
-12. Member of Fatigue Monitoring System Project Team. Following installation of the new system, a LCR will be reqaired to take credit for the more accurate fatigue usage factor.
Page 1 of 2
, o
e [ FOUR WEEK LOOK AHEAD KENNETH M. O'GARA l AUGUST 15, 1994 TASKS TO BE PERFOPRED fCont'd)
- 13. Address SERT 94 04 Open Item to resolve inadvertent SI at Power issue and PDP operation following SI with No loss of cffsite power. Need to SORC UFSAR change by end of september. Draft SE by M is currently under review.
- 14. Iesue revised Part 21 letter to NRC/on"eGMFIP microswi Jnes per comments from NSR '
COMPLETED TASKS FOR PRE"IOUS WEEE j 1. Resolved comments on HWCS letter. Issued for signature.
- 2. Rewrite of COMSIP PT21 report initiated
- 3. Resolved Ops comment on CCP LCR and issued for signoff
- 4. F042 letter to NRC issued for signature. Comments from HC Tech, Ops & Mech resolved To be completed this week.
1
- 5. LCR 93 10 letter issued for signoff. Resolved M.
Bur:: stein comments
- 6. -50.7 memo issued for Nuc. Depts, to roll down.
AREAS OF CJNCERN NONE MEETINGS AND TRAINING NONE VACATICN Planning to Take //22/94 as Vacation Page 2 of 2
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA AUGUST 29, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long I term commitments.-Draft _SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
. 2.- Complete SAR change, Salem /Nuc Dept. reorganization for i Salem and Hope Creek. Draft change and 50.59 Safety I evaluation with RAR for review and comn.ent by nuclear <
department.
- 3. LCR 92-07 for Hope Creek from GL 92-01 response._ Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR,
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address OA surveillance finding, minor TS change for the max. P-T curve heatup rate will also be addressed. Also, need to prepare letter to NRC requesting approval of Code Case-to extend P-T curves 10% during LTOP conditions. Presently, a draft letter to NME is under preparation to_ request resolution of comments on" calculation completed to support the use of RH3.
5, GL 92-01 update of HC UFSAR regarding'BWROG Equivalent Margin Analysis (1/31/95).- Submit justification for use of EOL USE values to NRC based on CEOG ef forts. (12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19 (09/30/94). With operations for signature.
7.- Revision to NAP-30 based on Biennial Review'(01/31/95) 8, 21.21(b) procedure changes and resolution.
- 9. Member of Fatigue Monitoring System Project Team.
Following installation of the new system, a LCR will be required to taks 'dit for the more accurate fatigue usage factor.
- 10. Address SERT 94-04 Open Item to resolve inadvertent SI at Power _ issue and PDP operation following SI with No loss of offsite power.
end of September. SE by Need to SORC UFSAR change by. H issued. NF to issue SAR change for review and comment & SORC the Change (10/31/94) Page 1 of 2 e e, l w -
I FOUR WEEK LOOK AHEAD f t
".ENNETH M. O'GARA AUGUST 39, 1994 Ij.9KL TO BE PERFORMED (Cent'd) '
- 11. Issue revised Part 21 letter to NRC on COMSIP microswitches per. comments from NSR. This< task is currently on hold'pending a final decision by Mgmt. on whether a change to the Letter is required.
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. HWCS letter Issued to NRC
- 2. Reviewed NAP-30 & NAP-57 combined procedure
- 3. Resolved Ops comment on CCP LCR.and issued for sign-off
- 4. -F042 letter to NRC issued
- 5. LCR 93-10 letter issued to NRC
- 6. Drafted a Letter NME & NES tb iguest resolution of comments on calculation completed to 2pport the use of RH3.
AREAS OF CONCERN POPS setpoint Issue Memo to NME & NES identifying cale concerns. Rewrite of COMSIP PT21 report (Drafted).-Is i*. Required? MEETINGS AND TRAINING NONE VACATIQ3 Planning to Take 9/19 - 21/94 as Vacation Page 2 of 2 f
- (
VI FOUR WEEK LOOK AHEAD KENNETH M. O'GARA SEPTEMBER 06, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued'for review. Need to complete 50.59 Safety Evaluation in support of change l (12/30/94)
! 2. Complete SAR change, Salem /Nuc Dept. reorganization for Falem and Hope Creek. l l Draft change and 50.59 Safety evaluation with RAR for review and comment by nuclear department. 1
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
4, DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address OA surveillance finding (due 09/30/94), minor TS change for the max. P-T curve heatup rate will also be addressed. This will need to be extended. Also, need to prepare letter to NRC requesting approval of code Case to extend P-T curves 10% during LTOP conditions. Presently, a draft letter to NME has been provided to DAS
-for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also, currently considering removinp the P-T curves from the Tech. Specs and incorperating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. The PTLR has several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values-to NRC based on CEOG efforts.(12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19 (09/30/94). With operations for signature.
- 7. Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision requests. Upon issuance, this task will be closed (01/31/95).
- 8. 21.21(b) procedure changes and resolution.
- 9. Member of Fatigue Monitoring System Project Team. Following installation of the new system, a LCR'will be required to take credit for the more accurate fatigue usage factor.
Page 1 of 2
'7 g g76C / >
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA SEPTEMBER 06, 1994 TASKS TO BE PERFORMED (Cont'd)
- 10. Address SERT 94-04 Open Item to resolve inadvertent SI at Power issue and PDP operation following SI with No loss of offsite power. Need to SORC UFSAR change by end of September. SE by H issued. NF to issue SAR change for review and comment & SORC the Change (10/31/94)-
- 11. Issue revised Part 21 letter to NRC on COMSIP microswitches per comments from NSR. This task is currently on hold pending a final decision by Mgmt, on whether a change to the Letter is required.
- 12. Provide response to NRC Request for Additional Information on HWCS. Issues include Tornado Missile analysis results for H2 and 02. Response due to NRC by 10/25/94.
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. NRC RAI on RWCS issued to NME to address comments. ATS I tasks assigned.
- 2. L&R Dept. Survey data evaluated for FXT.
- 3. Draft Letter to NME & NES issued to DAS for comment to 1
i request resolution of comments on calculation completed to support the use of RH3. 4. Reviewed draf t WOG report of Sample Tech Spec change s for implementing the Pressure / Temperature Limit Report for Salem. AREAS OF CONCERN POPS setpoint Issue Memo to NME & NES identifying calc Concerns.
- Rewrite of COMSIP PT21 report (Drafted). Is it kequired?
tiEETINGS AND TRAINING NONE VACATION c Planning to Take 9/19 - 2A/94 as Vacation
'/d cb afb;fM 5
Page 2 of 2
,/ '-
v
r ! V l FOUR WEEK LOOK AHEAD KENNETH M. O'GARA SEPTEMBER 26, 1994 l TASKS TO BE PERFORMED 1 l 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization for Salem and Hope Creek. Draft enange and 50.59 Safety evaluation with RAR for review and comment-by nuclear department.
- 3. LCR_92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address QA surveillance finding (due 09/30/94), minor TS change for the max. P-T curve heatup rate will also be addressed. This will need to be extended.
" Also, need to prepare letter to NRC requesting approval of Code Case to extend P-T-curves 10% during LTOP conditions.
Presently, a draft = letter to NME has been provided to DAS for review preparation to request resolution of comments on calculation completed to suppore the use of RH3. Also, currently considering removing i P-T curves from the Tech. Specs and incorporating t information into a Pressure / Temperature Limit Repoi 'TLR) for Salem. The PTLR has several benefits, and s.. auld be included as part of
-one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values to NRC based on CEOG ef forts. (12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19
, (09/30/94). With operations for signature.
7. Revision to_ NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided'on the draft to address the biennial review comments and procedure revision requests. Upon issuance, this task will be closed (01/31/95). 8, 21.21(b) procedure changes and resolution,
- 9. Member of Fatigue Monitoring System Project Team. Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor.
Page 1 of 2
, e
s
-FOUR WEEK LOOK AHEAD ;
KENNETH M. O'GARA SEPTEMBER 06, 1994 I
- i TASKS TO BE PERFORMED (Cont'd) l
- 10. Address SERT 94-04 Open' Item,to resolve inadvertent SI at Power issue-and PDP operation following SI with No less of offsite power. Need to SORC UFSAR change by end of September. SE by H issued. NF to issue SAR change for review and comment & SORC the Change (10/31/94) ;
l
- 11. Issue revised Part 21 letter to NRC on COMSIP microswitches per comments from NSR. This task is currently on hold pending a final decision by Mgmt, on whether a change to the Letter is required.
- 12. Provide response to NRC Request for Additional Information on HWCS. Issues include Tornado Missile analysis results-for H2 and O2. Response due to NRC by 10/25/94.
- 13. Finalize survey data for FXT based on Mgmt, input ,
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Second draft of Code Case Letter that includes latest cale results from NME completed (with,DAS). Need to issue for review and comment.
- 2. Draft Letter to NME & NES revised per meeting on 9/16. . Memo includes actions agreed to at this time by NME,NES and NLR.
l 3. Reviewed draft Prob. Report on POPS issue for NME. 4.- Reviewed SAR and Tech Specs. for W&S to address E&PB org responsibilities
.5,.
Drafted letter to QA requesting extension of Tech. Spec. Change on P/T heatup curves. (W/ DAS) AREAS ~OF CONCERN L
- Rewrite of COMSIP PT21 report (Drafted). Is it Required? ,
, MEETINGS AND TRAINING
- Operability Training 9/27
- W&S training 10/12 (12:30) & 10/13 (6:00) t VACATION-Was On Vacation 9/19, 20 and 1/2 day'9/21 6
l Page 2 of 2 _. - , , ,-v -r.yg. w g,.-
.w .,4---- ..,...y
i O' FOUR WEER LOOK. AHEAD KENNETH M. O'GARA-OCTOBER 03, 1994 i
.. TASKS TO BE PERFORMED
.1. '
Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review. Need to complete.50.59 Safety _ Evaluation in support of change (12/30/94)
- 2. ; Complete _SAR change, Salem /NucLDept.-reorganization-for Salem and Hope-Creek.- Draft change and 50.59. Safety evaluation with RAR for review and comment by nuclear-department.
3 ~. LCR 92-07-for Hope = Creek from GL 92-01 response. Tied to Capsule Removal in Springfl994 Outage. Current schedule is 4th Q 1994-for preparation'of LCR. 4. DEF regarding POPS.setpoint analysis. Need LCR to credit 'RH3 Relief Valve. To' address .OA surveillanca . finding- (due: 09/30/94),. minor TS change for the max. P-T curve heatup . rate will also-be addressed. This will need to be extended. Also, need to prepare. letter-to NRC requesting approval of Code Case to extend P-T-curves 10%'during LTOP conditions.-
-Presently, a draft letter to'NME has-been provided to DAS for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also,-
currently considering removing the P-T curves from the Tech.? Specs and incorporating this information into a Pressure / Temperature Limit' Report (PTLR) "for. Salem. The PTLR has several, benefits, and should be-included as part of
-one= Tech Spec change-to-address all of the issues above.
- 5. GL 32-01 update of HClUFSAR regarding BWROG Equivalent Margin (Analysis (1/31/95). . Submit, justification for tseso:
EOL USE values to NRC based' on CEOG. efforts 2(12/31/95)
;6.-
. Complete ChargingLPump/Boration Flow: Path LCR 93-19
,. (09/30/94). With operations for signature.
7.~ Revision to NAP-30 based on Biennial Review. As part of._the NAP improvement process, NAP-30 and 57 are being' combined into one procedure.- Comments were provided on the draft to: address.che biennial review comments and procedure revision requests. Upon issuance, this task will be closed (01/31/95). 86 21.21(b) procedure changes and resolution. !
- 9. . Member of Fatigue Monitoring System Project-Team.
Following
' installation ~of'the new system, a LCR will be required to
_ take credit for the more accurate-fatigue usage factor, m ; Page 1 of 2
- j i
m .~
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA OCTOBER 03, 1994 TASKS TO BE PERFORMED (Cont'd)
- 10. Address SERT 94-04 Open Item to resolve inadvertent SI at Power issue and PDP operation following SI with No loss of offsite power. Need to SORC UFSAR change by end of September. SE by H issued. NF to issue SAR change icr review and comment & SORC the Change (10/31/94)
- 11. Issue revised Part 21 letter to NRC on COMSIP microswitches per comments from NSR. This task is i
currently on hold pending a final decision by Mgmt. on whether a change to the Letter is required. j
- 12. Provide response to NRC Request for Additional Information on HWCS. Issues include Tornado Missile analysis results for H2 and 02. Response due to NRC by 10/25/94.
- 13. Finalize survey data for FXT based on Mgmt. input COMPLETED TASKS FOR PREVIOUS WEEK 1.
Draft of Code Case Letter that includes latest cale 7, 2. results from NME completed & issued for review and comment. Memo to NME & NES revised per meeting on 9/16 issued. l 3. Drafted letter to NRC or. HWCS. Input rec'd from NME. l- 4. Completed Peer review of CIV LCR. i 5. Memo to QA requesting extension of Tech. Spec. Change on P/T heatup curves issued
- 6. ' Addressed comments from J. Stone on LCR 93-10, Grid Voltage per telecon
- 7. Support provided_to QA Assessment re. Part 21
- 8. Began T/S change for PTLR AREAS OF CONCERN
~
- Rewrite of COMSIP PT21 report (Drafted), Is it Required?
MEETINGS AND TRAINING tW&S training 10/12 (12:30) & 10/13 (8:00) VACATION May take-10/11 and 10/12 off (Maybe only 1/2 day 10/12 so I can be here for W&S training) Page 2 of 2 i,.
'J "
FOUR WEEK LOOK AMEAD
-KENNETH M. O'GARA OCTOBER 17, 1994 TASKS TO BE PERFORME2
- 1. -Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety E.'aluation in support of change (12/30/94)
- 2. Complete SAR change, Salem /Nuc 9t. reorganization for Salem and Hope Creek. ' Draft aaa9e and 50.59 Safety evaluation with RAR for review and comment by nuclear department.
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address QA surveillance finding (due 09/30/94), minor TS ch: je for the max. P-T curve heatup rate will also be addressed. This will need to be extended.
Also, need to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions. Presently, a' draft letter to NME has been provided to DAS , for review preparation to request resolution of comments on ' calculation completed to support the use of RH3. Also, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. _The PTLR has several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GLL92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values to NRC based on CEOG efforts.(12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19 (05/30/94). With operations for signature.
- 7. Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision requests. Upon issuance, this task will be closed (01/31/95).
8, 21.21(b) procedure changes and resolution.
- 9. Member-of Fatigue Monitoring System Project Team. Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor.
Page 1 of 2
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA OCTOBER 17, 1994 TASKS TO BE PERFORMED (Cont'd)
- 10. Address SERT 94-04 Open Item to resolve inadvertent SI a at Power issue and PDP operation following SI with No- )
loss of offsite power. Need to SORC UFSAR change by; end of September. SE by H ise.74. NF to. issue SAR change- .! for review and comment & SORC tre Change -(10/31/94) l
-11. Issue revised Part 21 letter to NRC on COMSIP l microswitches per ceraments from NSR. This task is currently on hold pending' a final decision by Mgmt. on whether a change to the-Letter is requirad.
,12. Provide response to NRC Request for Additional Information on HWCS, Issues include Tornado Missile analysis results -
for H2 and 02. Response due to NRC by 10/25/94. COMPLETED TASKS FOR PREVIOUS WEEK (From 10/4 /94)
- 1. Resolved comments on Code Case Letter
- 2. Peer Review of-GL 88-08 reponse i
- 3. Letter to NRC on m4CS to.NES for signoff. Resolved DAS '
comment
- 4. Completed NLR survey for,FXT
- 5. T/S change for PTLR - T/S pages marked up and Sig. Haz.
initiated
- 6. CCP LCR approved & ready for SORC
- 7. Attended W&S training AREAS OF CONCERN
- Rewrite of COMSIP FT21 report (Drafted). IE it Required?
MEETINGS AND--TRAINING
- SORC on 10/19 VACATION Was off.10/11, 10/12 and 10/13 d
Page 2 of 2
/ r
e FOUR WEEK LOOK AHEAD KENNETH M. O'GARA OCTOBER 24, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94) l
- 2. Complete SAR change, Salem /Nuc Dept. reorganization for Salem and Hope Creek. Draft change and 50.59 Safety L
' evaluation department. with RAR for review and comment by nuclear
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address OA surveillance finding, minor TS change for the max. P-T curve heatup rate will also be addressed. Thic will need to be extended. Also, need'to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions (11/30/94). Presently, a draft letter te NME has been provided to DAS ; for review preparation to request resolution of comments on r calculation completed to support the use of Ri3. Also, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. The PTLR has several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values to NRC based on CEOG ef forts. (12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19 (11/15/94). Ready for SORC.
7 Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft-to address the biennial review comments and procedure revision requests. Upon issuance, this task will be closed (01/31/95). 8.- Member of Fatigue Monitoring System Project Team- Following installation of the new system, a LCR will be required to take-credit for the more accurate fatigue usage factor. - Page 1 of 2
4 ~ FOUR' WEEK LOOK AHEAD. KENNETH M. O'GARA-OCTOBER 24,- 1994
. TASKS TO'BE PERFORf!ED-(Cont'd)
- 9. Isrue revised Part'21? letter to NRC on-COMSIP microswitches per.: comments from NSR. This task is currently onthold pending a: final, decision b on whether a. change to the: Letter is required.:y Mgmt.
COMPLETED TASKS FOR^ PREVIOUS WEEK
- 1. Comments resolved and Ccae Case Letter provided to1NME fore sign-off~
2.- Reviewed:NFU memo to H on Inadvetent SI generic Implicatiens 3.-- Letter on HWCS'aubmitted~to NRC 10/24/94
- 4. Review of NOC commitments for;FXT ABJAS-OF CONCERN
~ Rewrite of COMSIP PT21 report- (Draf ted)'.;Is it Required?
r NOTE: WJM addressing;UFSAR Change onfInadvertent SI. -NFU to SORC-by'10/31/94.
~
- MEETINGS AND TRAINING
--SORC on 10/26 (Hopefully)
VACATION-
- None' Planned-e-
- . =
h
- r r -
Page 2 of 2 +.'
J* FOUR WEEK LOOY, AHEAD KENNETH M. O'GARA i
. OCTOBER 31, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term-commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
-2. Complete SAR change, salem /Nuc Dept. reorganization for Salem and Hope Creek. Draft change and 50.59 Safety evaluation with RAR for review and comment by nuclear department.
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th 0 1994 for preparation of LCR. ,
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 :
Relief Valve. To address QA surveillance finding, minor TS change for the max P-T curve heatup rate will also be 1 addressed. This will need to be extended. Also,~need to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions (11/30/94).
. Presently, a draft letter-to NME has been provided to DAS for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also, currently considering removing the P-T curves from the Tech.
. Specs and incorporating this information'into a.
Pressure / Temperature Limit Report (PTLR) for Salem. The PTLR has , several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of .
EOL- USE values to NRC based on CEOG efforts. (12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19
, (11/15/94). Ready for SORC.
- 7. Revision to NAP-30 based on Biennial Review. As part of the NAP. improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision requests. Upon issuance, this task ~will be closed (01/31/95).
- 8. Member of Fatigue Monitoring System' Project Team. Following ';
installation of.the new system, a LCR will be required to take credit for the more accurate fatigue usage factor-. Page 1 of 2
FOUR WEEK LOOK AMEAD
-KENNETH M. O'GARA OCTOBER 31, 1994 TASKS TO BE PERFORMED:(Cont'd)
- 9. -Issue revised Part 21 letter to NRC on COMSIP
-microswitches per comments from NSR. ~This. task is currently on hold pending a final decision by Mgmt, on i
whether a change-to the Letter is required. COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Completed InFpection Report review for SRM re. Potential SALP rating for this period. Completed IR aummarius.
- 2. Follow-up of GL 93-06,-Combustible Gases in Vital Areas, regarding NES response. DCP may be required.
- 3. Code Case N-514 Letter still with NME for sign-off.
AREAS OF CONCERN
- Rewrite of COMSIP PT21 report (Drafted). Is it Required? '
NOTE: NJM addressing UFSAR Change cn1 Inadvertent SI. -NFU to SORC by.10/31/94. MEETINGS AND TRAINING
- SORC on 11/02 (Hopefully)
VACATION-Took a Sick Day on Friday 10/27 i _2 _ i Page 2 of 2
V l FOUR WEEK LOOK AHEAD l KENNETH M. O'GARA NOVEMBER 14, 1994 TASKS TO BE PERFORMED l l
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review. Need.
to complete 50.59 Safety Evaluation in support of change (12/30/94) ,
-2. Complete SAR change, Salem /Nuc Dept. reorganization for Salem and Hope Creek. Draft change and 50.59 Safety evaluation with RAR for review and comment by nuclear department. ;
-3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address OA surveillance finding, minor TS change for the max. P-T curve heatup rate will also be addressed. This will need to be extended. Also, need to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions (11/30/94). -
Presently, a draft letter to NME has been provided to DAS for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. The PTLR has several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). -Submit justification for use of EOL USE values to NRC based on CEOG ef forts. (12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93 . (11/21/94).
7 Revision to NAP-30 based on Biennial Revi6kt. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision ! requests. Upon issuance, this task will be closed (01/31/95).
- 8. Member of Fatique Monitoring System Project Team. Following installation oc the new system, a LCR will be required to take credit for the more hecurate fatigue usage factor.
Page 1 of 2 o t
v. FOUR WEEK LOOK AHEAD KENNETH M. O'GARA : NOVEMBER 12, 1994 TASMS TO BE PERFORMED (Cont'd)
- 9. Issue revised Part 21 letter to NRC on COMSIP i
.microswitches per comments from NSR. This task is
-currently on hold pending a final decision by Mgmt. on whether a change to the Letter is required.
i
- 10. Complete Sustantial Safety Hazard Evaluation for Part 21 94-07, Feedwater Nozzle Flow Bypass by January 6, 1995, or submit Interim Report to NRC. NME input due to NLR by 12/19/94.
- 11. Complete Sustantial Safety Hazard Evaluation for Part 21 94-08, SW Pump Carbon Steel Retaining Rings by December 28, 1994, or submit Interim Report to NRC. NME input due to NLR by 12/12/94.
- 12. Complete Sustantial Safety Hazard Evaluation for Part 21 94-09, DG Fuel Pump Inlet Counterbore Area by January 3, 1995, or submit Interim Report to NRC. NME input due to NLR by 12/16/94.
- 13. Complete Sustantial Safety Hazard. Evaluation for Part 21 94-10, Hiller Actuator Failure by January 9, 1995, or submit In?.erim-Recort to NRC. NME input due to NLR by 12/23/94.
COMPLETED TASFS FOR PREVIOUS WEEK
- 1. Finalized NLR Client survey'for FXT.
- 2. LCR 93-19 SORC'd, ready for DAS signoff
- 3. Part 21 94-07 on Feedwater Nozzle, Memo to NME drafted
- 4. .Part 21 94-00 on SW Pump Retaining Rings, Memo to NME drafted.
- 5. Part 21 94-09 on DG Fuel Pump Inlet Counterbore, Memo co NME
. drafted.
- 6. Part 21 94-10 on Hiller Actuator Failures, Memo to NME drafted.
AREAS OF CONCERN
- Rewrite of COMSIP PT21 report (Drafted). Is it Required?
Code Case N-514 letter requires additional work by NME. NES (Bailey) won't-approve until issue .egarding PDP Operation has been dddressed by NME. Need to Df.scuss MEETINGS AND TRAINING ' Mtg. w/ J. Bailey on Code Case Letter. See Area of concern above VACATION , I would like to take the week of Thanksgiving off (11/21-11/25) Page 2 of 2 i
J FOUR WEEK LOOK AHEAD KENNETH M. O'GARA NOVEMBER 28, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
- 2. Complete SAR change, Salem /Nuc Dept. reorganization for Salem and Hope Creek. Draft change and 50.59 Safety evaluation department.
with RAR for review and comment by nuclear { L
- 3. LCR 92-07 for Hope Creek from GL 92-01 respons). Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4ta Q 1994 for preparation of LCR.
4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3
~
Relief Valve. To address QA surveillance finding, minor TS change for the max. P-T curve heatup rate will also be addressed. This will need to be extended. Also, need to l prepare letter to NRC requesting approval of Code Case to } extend P-T curv 4 10% during LTOP conditions (11/30/94). Presently, a draft letter to NME has been provided to DAS for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. The PTLR has several benefits, and should be included as part of one Tech Spec change.to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWRGG Equivalent Margin-Analysis (1/31/95). -Submit justification for use of EOL USE values to NRC based on CEOG ef forts. (12/31/95) 6, Complete Charging Pump /Boration Flow Path LCR 93-19
, (11/21/94).
7. Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision requests. Upon issuance,-this task will be closed (01/31/95).
- 3. Member of Fatigue Monitoring System Project Team. Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor.
Page 1 of 2
FOUR WEEK LOOKLAHEAD !
- KENNETH M.-O'GARA~
-NOVEMBER 28,-1994 i
TASRS TO BE PERFORMED (Cont'd) '?
.9. Issue-revised Part 21 letter =to NRC on COM3IP microswitches per comments from NSR._.This task is -
currently on hold pending a final- decision by Mgi..t. cn1 l whether-a change to the Letter is required. 10 . _ Complete _Sustantial Safety Hazard Evaluation-for Part-21 94-07, Feedwater Nozzle Flow Bypass by_ January 6, 1995,Jor submit Interim Report to NRC. NME. input due to NLR by 12/19/94;-
- 11. Complete Sustantial Safety Hazard Evaluation for Part 211 '
94-08, SW Pump Carbon Steel Retaining Rings by December 28,- , 1994, or submit Interim Report to NRC. NME input due to NLR I by 12/12/94. t 12 . - Complete Sustantial Safety Hazard Evaluation for Part 21 94-09, DG Fuel Pump Inlet Counterbore Area by January 3, j 1995, or submit Interim Report to NRC. NME input due to NLR ; by'12/16/94. *
- 13. _ Complete Sustantial Safety Hazard Evaluation for Part 21 94-10, Hiller Actuator Failure by January 9, 1995, or i
submit Interim Report to NRC. NME input due to NLR by 12/23/94. , COMPLETED TASKS FOR PREVIOUS WEEK LNONE - ON VACATION = AREAS OF-CONCERN
--Rewrite of COMSIP PT21 report (Draf ted)L. Is it. Required?
- MEETINGS AND TRAINING
" ~
-NONE' '
VACATION
; Week of Thanksgiving off__(11/21-11/25) '
i I Page-2 of 2 '
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA
. DECEMBER 12, 1994 TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term ccmmitments. Draft SAR change Issued for review. Need to c'mplete 50.59 Safety Evaluation in support of change I
(12/30/94) f 2. Complete SAR change, Salem /Nuc Dept. reorganization for l Salem and Hope Creek. Draft change and 50.59 Safety i evaluation with RAR for review and comment by nuclear department.
- 3. LCR 92-07 for Hooe Creek from GL 92-01 response. Tied to Capsule Removal 11-Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address QA surveillance finding, minor TS change for the max. P-T curve heatup rate will also be addressed. This will need to be extended. Also, need to prepare letter to NRC requesting approval of Code Case to extend P-T curves 10% during LTOP conditions (11/30/94).
Presently, a draft letter to NME has been provided to DAi for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem.- The PTLR has several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values to NRC based on CEOG efforts.(12/31/95)
- 6. Complete Charging Pump /Boration Flow Path LCR 93-19
. (11/21/94).
- 7. Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision requests. Upon issuance,_this task will be closed (01/31/95).
- 8. Member of Fatigue Monitoring System Project Team.
Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor. Page 1 of 2
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA DECEMBER 12, 1994 TASKS TO BE PERFORMED (Cont'd) 9. Issue revised Part 21 letter to NRC on COMSIP microswitches_per comments from NSR. This task is currently on hold pending a final decision by Mgmt. on whether a change to the Letter is required.
- 10. Complete Sustantial Safety Hazard Evaluation for Part 21 94-07, Feedwater Nozzle Flow Bypass by January 6, 1995, or l
' submit Interim Report to NRC. NME input due to NLR by 12/19/94.
- 11. Complete Sustantial Safety Hazard Evaluation for Part-21 94-08, SW Pump Carbon Steel Retaining Rings by December 28, 1994, or submit Interim Report to NRC.
- 12. Complete Sustantial Safety Hazard Evaluation for Part 21 94-09, DG Fuel Pump Inlet Counterbore Area by Janus',
1995, or submit Interim Report to NRC. NME input due to3,NLR by 12/16/94.
- 13. Complete Sustantial Safety Hazard Evaluation for Part 21 94-10, Hiller Actuator Failure by January 9, 1995, or submit Interim Report to NRC. NME input due to NLR by 12/22/94.
COMPLETED TASKS FOR PREVIOUS WEEK
- 1. Drafted LER response for POPS and_ resolved comments 2.
3. Received input for PT21 97-08, SW Pump rings - NR Drafted letter for FW nozzles PT21 94 Interim Report
- 4. Met with NRC to discuss POPS
- 5. Provided L&R input for NDRAP to add PDP pump trip
- 6. Code Case letter back with NME AREAS OF CONCERN
- Rewrite of COMSIP PT21 report (Drafted). Is it Required?
MEETINGS AND TRAIMING SORC Wed, for LER
. VACATION NONE I
Page 2 of 2
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA DECEMBER 19, 1994
-TASKS TO BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long term commitments. Draf t SAR change Issued for review Need to complete 50.59 Safety Evaluation in support of change (12/30/94)
- 2. Compinte SAR change, Salem /Nuc Dept. reorganization for Salem and-Hope Creek. Draft change and 50.59 Safety.
evaluation with RAR for review and comment by nuclear department.
- 3. LCR 92-07 for Hope Creek from GL 92 01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is 4th Q 1994 for preparation of LCR.
- 4. DEF regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address QA surveillance finding minor TS charge for the max. P-T curve heatup rate will also be addressed. This will need to be extended. Also, need to prepare letter to NRC requesting approval of Code Case to
' extend P-T curves 10% during LTOP condit'ons (11/30/94). Presently, a draf t letter to NME has beet, provided to DAS for review preparation to request resolution of comments on calculation completed to support the use of RH3. Also, currently considering removing the P-T curvas from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. The PTLR has
=several benefits, and should be included as part of c e Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values to NRC based on CEOG ef forts. (12/31/95)
- 6. Complete Charging' Pump /Boration Flow Path LJR 93-19 (11/21/94).
- 7. Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision requesta. Upon issuance, this task will be closed (01/31/95).
- 8. Member of Fatigue Monitoring System Project Team. Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor.
Page 1 of 2
l FOUR WEEK LOOK AHEAD
~- -KENNETH M. O'GARA DECEMBER 19, 1994 L
TASKS TO BE PERFORMED-(Cont'd)
- 9. ' Issue revised Part 21' letter to NRC-on.COMSIP microswitches per comments from NSR. This task in l
- currently on hold pending a_ final _ decision _by-Mgmt. on:
'whether a change to-the Letter is1 required.
L
- 10 . . Complete Substantial-Safety _ Hazard Evaluation-for Part'21 94-07, Feedwater Nozzle-Flow-Bypass by January 6,'1995,. or j
i submit: Interim l Report to NRC.- NME input due to NLR by_ " 12/19/94.
- 11. Complete Substantial Safety Hazard; Evaluation for Part 21 94-08,- SW Pump-Carbon Steel Retaining Rings by December 28, 1994, or' submit Interim Report to NRC.
- 12. LComplete Substantial Safety Hazard Evaluation for:Part 21 94-09, DG Fuel Pump Inlet Counterbore Area by-January 3, 1995,.or submit Interim. Report to NRC. NME input due to NLR by 12/16/94.
- 13. Complete Substantial; Safety Hazard Evaluation for Part'21' 94-10, Hiller Actuator Failure:by January 9, 1995, or:
submit Interim Report to;NRC. NME input due: to NLR by: 12/23/94.
- COMPLETED TASKS FOR PREVIOUS MggE 1.:SORC'dLLER ret) nse for POPS 2.iIssued PT21 97'08, SW Pump rings to NME for signoff'
~3.:-NRC letter for-FW nozzles PT211- 94-07.'-_ Interim Report for
-signoff,1 comments resolved 4.--Issued = letter to NRC on-POPS Admin Controls-
- 5 ~. Issut? PT21: 97-09,E DG counterbore: to -NME -for'signof f 6.' Code Case letteriin sign-off
, AREAS OF CONCERN
- Rewrite'of COMSIP PT21 report (Drafted). Is'it Required?
MEETINGS AND TRAINING SORC Wed. for K^.59 on POPo?T/S Bases-
- VACATION
^
Out: Monday & Tuesday after Christmas (Holidays)
~
Page.2'of 2
i FOUR WEEK LOOK AHEAD-KENNETH M. O'GARA JANUARY ~23, 1995 TASKS TO BE PERFORMER '
- l. Initiate SAR change for incorporating Salem ATWS long term commitments. Draft SAR change Issued for review.fNeed '
.to complete 50.59 Safety Evaluation in support of change (12/30/94) 2.. Complete SAR change, Salem /Nuc Dept. reorganizat.on for d Salem and Hope Creek. Draft change and 50.59 Safety j evaluation with RAR for review and comment'by nuclear i department.
]
3; LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage. Current schedule is i 4th Q 1994 for preparation of LCR.
- 4. DEF/LER regarding POPS setpnint analysis. Need LCR to credit RH3 Relief Valve. To address QA surveillance finding, minor I TS-change for the~ max. P-T curve heatup rate will also be addressed. This will need to be extended. Letter to NRC requesting. approval of Code Case to extend P-T curves 10%
during LTOP conditions issued-12/22/94. Also, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Sa.em. The PTLR has several benefits, and should be included as part of'one Tech Spec change-to address all of the issues above.
-5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent Margin Analysis (1/31/95). Submit justification for use of EOL USE values 'to NRC based- on- CEOG efforts. (12/31/95)
- 6. Revision to NAP-30 based on' Biennial Review, fAs part of the
~
NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision
, requests. .Upon issuance, this task will be closed (01/31/95).
- 7. . Member of Fatigue Monitoring System Project Team. Following
' installation of the new system, a LCR will be required to take credit for thn moi = accurate fatigue usage factor.
l Page 1 of 2
t i
.j
. FOUR WEEK LOOK AHE@ >
KENNETH M.-O'GARA' '
. January 23, 1995;; ;
' TASKS TO BE PERFORMED (Cont'df
; I
.8.- Issue revised Part 21-letter'to NRC on.COMSIP- 1
--microswitches per comments from'NSR. .=This task-is currently'on-hold pending-a finalDdecision by-Mgmt. on. I whether a-change to the. Letter is-required. ;
-t 19 . - Issue'1etter to-NRC-regarding use of, alloy 300 steam .'
generator tube plugs'(1/31/95), _ i
- 10.
~
Support enforcement conf. on 50.7' issue. t COMPLETED TASKS-FOR-PREVIOUS WEEK
-1. . Letter'on SG tube plugs'w/ DAS. .;
2.-Prepped Enforcement conf. slides _ majority of week.
'r
- 3. Prepared. chronology on POPS issue for FXT, AREAS OF CONCERN I
- Rewrite'~of COMSIP'PT21' report (Drafted). Is it. Required? :
i, MEETINGS'AND TRAINING Completed. GET re-qual . '(100%) . . I Numerous 1 meetings setup for Enf. Conf.1 Dry Runs. . VACATION NONE. 2
._. +
+
Page 2/of 2 4.
.s ,
, , .a.m ., , , .,,_.J,,.._ ..,.f .
,, ,, _, _4,,_ , . . . . . _ , , , _ , - , . . , , _ . , _ , _ , , _ . _ _ , , , . . , .,
I FOUR WEEK LOOK AHEAD l KENNETH M. O'GARA JANUARY 30, 1995 TASKS TO-BE PERFORMED
- 1. Initiate SAR change for incorporating Salem ATWS long (
term commitments. Draft SAR change Issued for review. Need to complete 50.59 Safety Evaluation in support of change - IO3/30/95) 0
- 2. Complete SAR change, Salem /Nuc Dept, reorganization for Salem and Hope Creek. Draft change and 50.59 Safety evaluation with :'.AR for review and comment by nuclear department.
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to- ?
o Capsule Removal in Spring 1994 Outage. Current schedule is 4th O 1994 for preparation of LCR.-
- 4. DEF/LER regarding POPS setpoint analysis. Need LCR to credit RH3 Relief Valve. To address OA surveillance finding, rinor TS change for the max. P-T curve heatup rate will also be addressed. This will need to be extended. Letter to NRC requesting approval of Code Case to extend P-T curves 10%
during-LTOP conditions issued 12/22/94. Alsc, currently considering removing the P-T curves from the Tech. Specs and incorporating this information into a Pressure / Temperature Limit Report (PTLR) for Salem. The.PTLR has several benefits, and should be included as part of one Tech Spec change to address all of the issues above.
- 5. GL 92-01 update of HC UFSAR regarding BWROG Equivalent e Margin Analysis. Submit justification for rse of n@
EOL USE values to NRC based- on CEOG efforts. (12/31/95) 6, Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined into one procedure. Comments were provided on the draft to address the biennial review comments and procedure revision
, requests.- Upon issuance, this task will.be closed (02/28/95).
- 7. Member of Fatigue Monitoring. System Project Team. Following installation of the new system, a LCR will be required to
,take credit for the more accurate fatigue usage factor.
a Page 1 of 2
c-s FOUR. WEEK LOOK AHEAD KENNETH M. O'GARA January 30,'1995
. TASKS TO BE PERFORMED (Cont 8d)
{ 8 .- -Issue: revised'Part 21; letter.to NRC on COMSIP, i microswitches per comments from NSR.- This task is currently on hold pending-a _ final decisjon-by Mgmt. on J
.whether a-change to_the_ Letter is required.
'9. - : Complete Substantial: Safety Hazard eval-of Part 21, 95 .
Anchor Darling check valves by -(3/27/95) . 10 . ' Complete Substantial Safety Hazard eval of Part-21, 95 Diesel' Generator Turbo-Luost Air. Controller (3/27/95)~.- -! 11.- Support enforcement conf. on 50.7 issue..
- 12. Alloy 600 SG tube plug letter-to NRC requires a letter to l nrc within 30 days of outage with any changes to Action Flan forfrepair-replacement. Awaiting issuance of Addendum 3 of l WCAP-12244 before a. schedule can be-developed. ID4E.toistatus 1 on 2/28/95.
COMPLETED TASKS FOR PREVIOUS WEEl$ ( _. I 1.. Resolved SLD'comm a on SG tube plugs letter and-issuedito, u
'.NRC on 01/30/95 2.. Supported 50,7'E forcement conf.-majority of weer.
- 3. Issued. chronology on POPS issue to NSR/
- 4. Drafted Part- 21 95-01, cA D Check Valves,1 memo to NME - w/ DAS,
- 5. Reviewed Part 21'95-02, DG Air Controller,-will' prepare SSH' eval and close. Based on IR, not reportable. '
. AREAS-OF CONCERN NONE.
. MEETINGS AND TRAINING
- NONE
. VACATION-NONE iPage 2 of 2 e ,
g.
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA MARCH 6, 1995-TASKS TO B2 PERFORM 1Q 1. Initiate SAR change for incorporating Salem ATWS long term commitmants. Jraft SAR change Issued for review. Need { to complete 50.59 Safety Evaluation in support of change (
=(03/30/95) I
- 2. Complete SARLehange, Salem /Nuc Dept._ reorganization-for Salem and Hope Creek. Draft change and 50.59 Safety evaluation with RAR.for review and comment by nuclear department.
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to Capsule Removal in Spring 1994 Outage.
4th'Q 1994 for preparation of LCR. NeedCurrent schedule is to submit summary report to NRC in accordance with APP. H by 4 /13 / 95. ! 4. DEF/LER regarding POPS setpoint analysis.-Need LCR to credit RH3 Relief Valve. To address QA surveillance finding, minor i TS change for the max. P-T curve'heatup rate will also be l' addressed. This will need to be extended. Letter to NRC requesting approval of Code Case _to extend P-T curves 10% during LTOP conditions was approvedon 2/13/95. Unit 1 Bases change prepared and issued for review and comment - Also.
-currently considering. removing the P-T curves from the Tech.
Specs and. incorporating this information into a Pressure /Tet.iperature Limit' Report (PTLR) for Salem. The PTLR has several benefits, and should be included as part of one_ Tech Spec change to address =all of the issues above. 5, GL 92-01 update of HC UFSAR regarding-BWROG Equivalent Margin Analysis. Submit justification'for use of EOL USE va'.ues to NRC based on CE03 ef forts. (12/31/95)
- 6. Revision-to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and 57 are being combined
. into'une procedure. ~
Comments were provided on the draft to address the biennial review comments and procedure revision requests. Upon issuance, this task will be closed (02/28/95).
- 7. Member of Fatigue Monitoring. System Project Team. Following installation of the new system, a ,CR will be required to take credit for the more accurate fatigue usage factor.
Page 1 of 2 I o ~
~
u n FOUR-WEEK 3.OOK-AHEAD-KENNETH M.:O'GARA-MARCH 6, 1995-1 TASKS-TO BE'-PERFORMED (Cent'd)- [8. Issueirevised'Part 21 letter-to:NRClon COMSIP.
. 4
'microswitches per comments-:from-NSR.- This task is
' currently on. hold pending a; final decision by Mgmt.--on Lwhether_a change to the Letter is: required;- )
9 --.
--Complete Substantial.-Safety Hazard eval'of Part 21,n95-01 --
g ; Anchorf Darling; check valves bye (3/27/95)-. - hm.,n.t , w J. 5 0. ~ amuV _10.1 Alloy 600 SG tube _ plug letter to;NRC; requires-a. letter to , 1 ' NRO withinl30 days:of outage with any.changestte Action Plani for repair-replacement. Addendum 3 of.WCAP-12244'was ) J reviewed 1by NMElon 2/23/95. For. Salem Unit-1,. no--impact on
~
the hot-leg tube' plugs and the-~ earliest' action for the cold
~ leg plugs is--in-tho year 2023'. For Salem Unit 2, 441 hot'. leg-L l ~ '
- plugs will require;; replacement _during theLnext. outage. This
-task has;been assigned to NME<to status plans for outage.on 9/1/95.-
R COMPLETED TASKS'FOR~ PREVIOUS WEEK'
~ ~
1.JCompleted braft of Bases..C'hange on POPS for-' Unit 1. Outifor' -i
. comment.
2.-Dratted and addressed' comments on. Code' Case memo for FXT.
;3. Completed database search of_.BWR' Code Repairs for!W.-Maher.
- 4. Followup on' capsule summary _ report submittal'.to NRC.
;3-Meme AREAS'OF b RT cn CONCERN Itpp%=A q Glco % rc sMM {d4 AWC
. g" -ScheduleLfor to-3/10/95., NMElto complete Part 21.95-0001 evaluation slipped .
m EETINGS AND-TRAINING-JLAWYERS-g ' VACATION Last week of March q ' (27-l31)3 -- q 4 .3 a
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FOUR WEEK LOOK AHEAD KENNETH'M. O'GARA
-MARCH-13, 1995-TASKS ~TO BE PERFORMED
- 1. Initiate SAR change for inec._ :m .p3 ealem ATWS long term commitments. Draft SAR c *a Int ~d for review. NeedL to . complete 50.59 Safety Eval 2 9 -
support of change. (03/30/95)
- 2. Complete SAR change,--Salem /Nuc Dep.. rganization for Salem and Hope Creek; Draft change and 50.59 Safety evaluation with RAR;for-review and comment by nuclear department._ '
-l
- 3. LCR 92-07 for Hope Creek from GL 92-01 response. Tied to-Capsule Removal in Spring 199410utage.- Current cchedulelis-4th Qu1994 for preparation of LCR.- Need to-submit summary L j
- r. report to NRC in accordance with APP. H by 4/13/95. '
f' 4. DEF/LER regardin. POPS setpoint-analysis.-Need LCR to credit RH3-Relief Valve. To address'QA surveillance rinding,. minor-TS change for the max. P-T curve heatup. rate will-also be addressed. This will need to be extended. Letter-to NRC '- requesting approval of Code Case to extend P-T curves 10% _! during LTOP conditions was approvedon_2/13/95.-_ Unit 1_ Bases j change prepared and issued for review =and comment. 'Also,. currently 1considering removing the-P-T curves from the Tech. 3 J Specs and it corporating this information into a Pressure /Ter. erature Limit Report .(PTLR) for Salem' The PTLR-has several benefits, and should be included as part of ' one Tech Spec change to address ~all-of the issues above. 5.- - GL 92-01' update of HC UFSAR'regarding BWROG Equivalent Margin Analysis. ' Submit justification for use of EOL-USEuvalues to NRC-based on CEOG efforts.(12/31/95)
- 6. Revision to NAP-30 based on Biennial Review. As part of the NAP improvement process, NAP-30 and:57 are-being combined
. into-one procedure. Comments were provided on the draft to address-the biennial review comments and procedure revision requests. Upon issuance, this task will be closed-(02/28/95).
7. Member of Fatigue' Monitoring System Project Team. Following installation of the new system, a LCR will be required to take credit for the more accurate fatigue usage factor. Page 1 of 2 [, f.0
FOUR WEEK LOOK AHEAD KENNETH M. O'GARA MARCH 13, 1995 TASKS TO BE PERFORMED (Cont'd)
- 8. Issue revised Part 21 letter to NRC on COMSIP-microswitches per comments from NSR. This. task is.
currently on hold pending a final decision by Mgmt.;on whether a change to the Letter is required.
- 9. Complete Substantial Safety Hazard eval of Part 21,_95 Anchor Darling check valves by (3/27/95).-
- 10. Alloy 600 SG tube plug letter to-NRC requires a letter to NRC within 30 days of outage with any changes to Action Plan for repair replacement. Addendum 3 of WCAP-12244-was reviewed by NME on 2/23/95. ~For Salem Unit 1,-no impact on the hot leg tube plugs and the' earliest action for the cold' leg plugs'is in the year 2023. For Salem Unit 2, 44' hot leg plugs will require replacement during the next outage. This task has been assigned to NME to-status plans-for' outage on 9/1/95.
COMPLETED TASKS FOR PREVIOUS WEEK 1. Completed draft of Bases Change on POPS for_ Unit l'. Out for' i comment.
- 2. Drafted Part 21 closeout memo. Need ME memo.
- 3. Drafted Specimen Capsule Report Letter to NRC.
- 4. Drafted LCR on HC P-T limit curves AREAS OF CONCERN Schedule for NME to complete Part 21 95-0001 evaluation slipped to-3/10/95. Still have not provided input.
MEETINGS AND TRAINING LAWYERS VACATION Last week of March (3/24-4/3) C I Page 2 of 2
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0 WINSTON & STRAWN , l 35 Was? WACKR DRIVE 1400 L STREET, N.W. 175 WATER FTREET CHICAOO ILLINOl5 606019703 WAllGNGTON, DC 20005 3502 NEW YORK, NY 10(DM61 002)371 5100
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DATE: September 05, 1995 l PLEASE DELIVER THE FOLLOWING PAGE(S) TO: Keith Lo, . FIRM: U. S. Nuclear Regulatory Commission FAXH. 916103375131 FROM: Marcia R. Gelman PHONE: 202-371-5747 ,
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- PROJECT.XLS "i " ~
, .e Abood, John- -Nancei,eae Amiemeissier Alliegte, Gine secroiery -
AmicuCCi, Joseph supervoet, cp.:eele Anhwere Alexandra Asseeme Ewaset ' teny, Dartisi - it.eed Eaease' tentens, Reben .ooe eas-Blair, John Lees tap **w Siecidev, Serbars - pieni cart - Stanch, Vie ' oes Eaiev cieet A ' Caldwell, Noel' se, steet tapnew Cauchnicki, Eugene- se si.et Enease, D'Sousa, Kewm -se. seen Eapae < Day. Sharon Aesensereeve cien F_edden. Thrams cem,awas nnwee cime Fortenberry, John mes.or fuears sweerswee Geisyst, Salvador Lead esueleer lasteusio#4r nrwiese Galleshow, Peter P enes aaeasem 6,liespie, Scott Paaesel safety Rev. Eas.
.4' Glynn, Mary _
E: _ -; as.eieni Gordon, John. llavenewy Aenesaw Graham, Thomas sr. siert sveiem. Aa see Gupta, Verendra : ' 19 es, Presset Eapanee Hartic'k, Jarnes scene Eaenew - Motate. Dale Assemens Eneaser 64udsen, Civet ' Teenned Asee(ees _ r4udson Cynthis - Ae mesesows cieet lily, Lassic onelio safest Reumow Ene Keen, Cheryl Teesumeas aseeelese Kirkpatrick, Wuaam ss. seussee, easiamwisve, Lark, John sesent seen Eaenow ~ l Laeksti. Chandre swear steet Ensiasee Lee, Oliva - LeesEnoese, Leonardi, Paul Aeuwesereew Liden, Edwin A. Preseen luier.4:estee:4 leaves ' _ _ _ _ Lloyd, Jeanne - Cweesowesene seeeetsey - Lodos, William Assestate Easlaew i Maldonando, Jorge - LeadEaeaew ' naecavoy. asament leusseer easia eeeeh Ochs, Marjorie A -- Geomesue ooewaisees& neeers - Ounden George Lees aseieaar Owen, Glen enaespas easinese Parmelee, Alan. Lees Empase' Plotti, Leonerd - sr. siew Eaginaw Plymate, Diene Tom see,mene Roon, Estela swene swt Eaenas. Retter, Lawrence A. Die.. Peessestenevowse mae Ritter, Semuel teussear amonceasses s,w Sangens, Nareinh s. w sieffE g nem l Sevrin, Norman .s da sten Ep l Sowak, Nicholas lLeadcomease.eeucaea, Page1
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PROJEC s.XLS v w.m. . rc3 i _ Ssmmerman, John _o ,eser Smith, Stanley Pnncies siet6ne Amore.ies Stomen. Dou0tas s.n,. st.n Enon Thomas. 8. David se. si tf taen e Tramontana. Edward se. stoff systems Am.6voi Troutman.- James - Ice.i a sowee a.ed. c Vales. Rocatio 'E ng c.or i Wright, Lorlie 'D..ument heceede c14 l Young, Getand mene.no asch wo imo ca.re A b Page 2 l
-__ - -- - - - - - - - - - - - ~
h V D . N M EVALUATION OF NOW-CONSERYATIVE LOW TEMPERATURE Uvr.mPRE85URE
~
SFTPOINT FOR SALEM >1 AND 2 BACYCROUND INPO Operating Experience #5832 from COMMED and LER 92-19 from Houston Lighting Overpressure andsetpoint system Power indicate that the low temperature (POPS system at Salem) may be non-conservative because the pressure drop due to instrument location was not considered'by Westinghouse in their analysis.
' " heat input transients and date.raines/ validates setpoints for th PORVsjthat will prevent the plant from exceeding its 10CFR Part 50 Appendix G curve,during the assumed transient. The Appendix G curves are based on a fracture mechanics approach to brittle fracture which beltline. assumes that a limiting flaw exits at the vessel At Salem the POPS system utilizes wide range RCS pressure psig. sensors PT403 and PT405 to actuate the PORVs at 375 The pressure sensors PT403 and PT405 are located on RHR suction line connected to the 11RC loop hot let. Westinghouse did not consider the dynamic pressure difference between the downcomer *(vessel beltline) and hot leg (which varies depending on the number of RCP in operation) and the static pressure difference between vessel beltline anl hot leg. These discrepancies result in s higher pressure at the vessel beltline than that the pressure at the sensing point,which leads to the concern the POPS setpoint may not adequately protect against exceeding Appendix G limit.
The Salem POPS analysis was done in-house using methodology provided by Westinghouse in their report "Prassure Mitigating Systems Transient Analysis" (July 1917 . Westingnouse has confirmed the methodology in thisN Tso)neglectedShressure difference of concern. ' EVALUATION of 74 osi In a telecon pressure Westinghouse difference ef "' provided a generic value,for the total operating j psi for a 4 loop plant with all RCPs The Salem POPS analysis (SGS/M-DM-042 and 062) calculates a maximumpressure
-maximum overshoot of 71 psi (mass input case). Therefore the reached at the pressure difference between thethebeltline vessel beltline correcting for and hot leg is 375 +
74 + 71 = 520 psig. Comparing this pressure to the composite Appendix G curve shows that the Appendix G limit would be exceeded below 180*F on Unit 1 and 135'T on Unit 2. (The composite curve is a combination oft te heatup curve and the 0*F/hr cooldown curve jper Westinghouse recommendation.) 4\sh
1; .
~
_W The POPE analysis 6e SGS/M-DM-042' and 052 assumed ac'tuation of 1
~
PORV the-otheri PORV being the single failure. In March 1988 W
-_ estinghouse-issued WCAP-11640.* Cold Overpressure Mitigation -
system Deletion Report", which provides criteria and methodology
- - for taking credit for the RHRyrelief valve to mitigate low L ', - temperature overpressure-events.- Here again the pressura difference of concern was--neglected and must be added to the--
70-ji: results_obtained using the methodology.
.e -a
[ One~of the requirements-for taking credit for the relief valve is
'} the deletion of the Auto Closure Interlock on'the'RHR suction
-isolation valves 1RH1'and 1RH2.
(1 h implemented at both-Salem Units. This deletion of ACI has been The-RHR relief valve has a setpoint of 375 psig which is the same as the POPS PORV setpoin't. ff 3. The relief valve being a spring loaded valve can' respond more quickly to a pressure transient. 4:
'M .An evaluation was-performed of the mitigating capability of the 3
RHR relief valve using the criteria and methodology of WCAp 11640 (See' Attachment). The results.show that if credit _is taken for
-[ i the' RHR relief valve the peak pressure alt the valve i.e. hot leg ML twill,not exceed 412.5 psig (1104 of' set pressure).- Adding tha 74 0 ._ _ psi ^ pressure difference gives-a peak pressure of 486.5 psig at ~! I the-beltline. The RHR relief valvp canLbe taken: credit for below - '1 RCS: temperature of 150*F. A-u s=
ke s AsA < temesta Err N rw- W4 cve a %fg ,j mCAac.tu and ddd.bm/ j;$wenA, wek is eeded 6ebre I cid<l cm je W g, Qa,ta. , at-temperatures greater than 150*F...there is i n l s_ increasing margin between the POPS setpoint and Appendix G curvea?d% RP. Pows would provide adequate mitigation an/ /4 A W AS W *
- ui dra f not' hs s W k 1Ah*~ CrtM & ls mia s=4~ 4 rcP W"T* **'d j
The loweston Therefore pressure.on Unit 2'the Salem Unit 2 Appendix G curve is'500 psig. Appendix G-limits are not exceeded during low temperature-overpressure transients, with mitigation provided-
,by RHR. relief valve below RCS temperature of 150*F.
F The lowest pressure on-Salem Unit 1 Appendix G curve is 450 psig.
- Q' H g psig below on Therefore Unit 1 the Appendix-G limit will be exceeded by 36 RCS temperature'of 170*F. Per an ASME code case on h" LTOP orotecti g "LTOP systems chall. limit the maximum pressure in lQ4 fthe /. vessel tb 110% of the presst.re determined to satisfy Appendix f
G Unit of Section XI Article G-2215". Invoking this code? case for 3 Li 1. increases the lowest pressure on Appendix G curve by 10% 9 d to 495 psig, which,is greater then the peak pressure of 486.5 J i using RHR relief' valve._~Therefore on Unit 1, 110% of Appendix G limit is met using RER relief valve for'RCS temperature below
" f_ 150*F and POPS above 150*F.
- f
- f. By taking credit for RHR relief valve restrictions on RCP operation can be avoided.
As discussed in the Attachment the RHR relief valve meets the criteria for use as a overpressure mitigation system. _,,_l-......, '
b eson $I 1. I T' CONCLDETOKS The POPS and the RNR relief, valve together mitigation
.for Galen Unit for_l'and design 2. low tamparature overpressur provide ~ adequate e transients, a
Initiator - Date
~
Reviewer - Date
~
~ supervisor ~ ,, '" A ge , g
\
Y. Date- $ H-
. en L .
i
n. t . ATTACENENT EVALUATION OF LTOP CAPABILITY OF RER RELIEF VALVE < W WCAP 11640 " Cold overpressure Mitigation-System Deletion Report", provides the methodology to justify using RHR relief valve for LTOP mitigation. Per WCAP 11640 if the Autoclosul.. Interlock on RHR suction valve is removed, which is the case at salem, then inadvertent isolation of RHRS relief valve is highly unlikely and the relief valve is available for LTOP mitigation. CRITERIA TO BE MET (SECTION 13 OF WCAP 116401
- 1. Avoid overpressurization of RHR. System above the valve accumulation pressure (maximum of 110% of valve setpoint) for design basis transients.
- 2. Prevent the pressure excursion,-that results from the design basis transients from exceeding the Appendix G limit. By invoking the LTOP ASME Code case the allowable limit is 110%
i of Appendix G limit. IV&EUATION la LTOP design transients fro Salem are: ggy _ iwt (Tech Spec 3.4.10.2 and-3.5.3 and Bases) 133 0%N Start of an idle RCP with the secondary water temperature of g y .i 1. SG less than or equal to 50'F above the RCS cold leg
;Qg:
j3 jp. temperature.
;\ D ' The start of a safety injection pump and injection into a
}. ) c water: solid RCS. The runout flow is 675 gpm based on LCR 91-03 issued to relax ECCS Operating limits.
< ' \$ "
,- . A
-The RHR relief valve capacity is 900 gpu at setpoi; t of 375 y
etl *[%} . psig + 10% accumulation per SGS/M-DM-042. N J\ Per WCAP 11640 a 470 gpm relief capacity would be adequate t-F0
; hX maintain peak RCS pressure below valve accumulation pressur_ for 1 ff . ,t the heat input transient initiated from a RCS temperatg e as high 5*
s t as 150'F (i.e. SG at 200'F). The 470 gpm size i$ PE3sa en
- gy transient being initiated from SS(t,emperature no greater than p g) 200'T (RCS at 150'F). At higher temperatures the required capacity woz6LA Therefore credit will be taken for RER .
l'gsi g g ' relief valve only below 150'F. h'v8- 154'/ Me r' i.5 ar tEce JNPchs
. Mar 6w
- a,
% udDe4 ~ L Pct.s serp.# and 14 NpwGt & I.e r and k e n Pers adue I $ With respect to the mass input transient the runout flow of I
_4 g.y@ -Safety Injection pump (design basis transient) is less than RHR y3 q -j relief valve capacity and can be handled by the relief valve. j
'* h,
i .- The peak pressure possible will be no greater than the
. accumulation pressure-and.therefore criterion:1 is set.-
Conservatively'the peak pressure'can be assumed to-be 110% of_
. valve set pressure when_ is Llox. 375 = 412.5 psig. Adding the pressure line_of_486.5 psig.
difference of 74 psi gives a peak pressure at the belt-The peak-pressure of 486.5 psig does not exceed the composite Appendix G limit fro Unit 2 and 110%_of Appendix G limit for Unit 1 and criterion 2 is met. By taking credit for RHR relief valve only below 150'F the issues related to valve and discharge piping capabilities under two-phase conditions and backpressure on the valve are precluded (Ref PSE 90-540).
~
Initiator Date ,.
.i -
.gl;"
Reviewer Date Supervinc Date- -
.. ,s- <x ..
N SC $,URE --
%ED QISJRIBUUCN -]/ FORQS[
INVESTIGATION STATUS RECORD w Case No.:- 1-95-013 Facility: SALEM UNIT 1
= Allegation No.:? RI-94-A-0159_ LCase Agent:_ LOGAN Docket No.: 50-272 Date Opened: 02/21/95 l Source-of Allegation: ALLEGER (A)
Notified by:;18. McDERM0TT '(DRS) Priority: HIGH (Coordinated with T. Nartin,RA) Category: :AE- - Case Code:: - RP (Power Reactor) Subject / Allegation: .P0TENTIAL INTENTIONAL OPERATION OF PLANT OUTSIDE ITS DESIGN BASIS AND FAILURE TO EVALUATE AN UNRESOLVED SAFETY-QUESTION'
- Remarks:- '
Monthly Status-Report:- 01/21195: On August 8,1995, the NRC received an allegation identifying 23 separate-concerns with-Salen operations, engineering, and: management.- The alleger was initially interviewed by Region'I' staff. 'On September 30, 1995, 01 initiated an, investigation into-the alleger's claims of. harassment,1 intimidation, and discrimination (1-94-043).- The. alleger was interviewed by 01 on: December ~'15, 1994. 10ne of the safety issues reportedly raised by theTalleger concerned
- the- plant's pressurizer overpressure protection system (POPS). 'It
- appears from the preliminary results of-an inspection by the-
. regional staff .that on or: about Apri1L 19,1994,- PSE&G changed the q
> POPS design basis transient for mass addition without evaluating the
-change pursuant to 10-CFR 50.59. PSE&G'did not make a-10 CFR 50.72. -
report to the NRC on this issue '(i.e.,Junanalyzed ondition) until: ! November 1994.
;0n February 16 -1995 an enforcement panel-convened to discuss =the
-alleger's concerns, s,pecifically whether the' licensee " failed to
= notify the NRC for 11 months & continued to operate-(the) plant cutside (its). design basis for POPS:(50.72)." LIn the 01 interview, the alleger stated that he brought to-the attention of.his supervisor information regarding.the POPS issue in the-form of an Incident Report (IR)-dated January 31, 1994. -He stated that.when he wanted to bring the IR to the control room, he was' told, "No. You do not file this. It's not your_ job . . ." Because this matter is viewed as a high priority by the Regional Administrator, and involves a distinct issue of alleged intentional wrongdoing by the licensee, 01 is opening a separate investigation to addru s this matter, and will coordinate with continuing regional inspection f fforts. Status: -FWP- ECD: 09/95.
; -p\
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7 j iLIMI,TEDsDJSTRIBUTION -- NOT FOR PUBLIC DISCLOSURE WITHOUT 01 APPROVAL ( v y -,
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'PuDlic Service Dectric and Gas Comosny - P.O Box 236 Hancock $ Bridge , Ne ec. *^~ -" NS_3$,,_
. Salem Generating Station . j L;': . 4
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De ember.14,mI,9,71,j. . .
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U. S. Nuclear' Regulatory Commission
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. Document control' Desk I T i-:--- p 4.2 1 Washington,-DC _20555
" ff; -42 L. . L /.M:j' i _,
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Attn.: Document CentrolLDesk
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SALEM GENERATING STATION-d .-6. - @ *"..:. t~ --- u . .;-- ... . 2.;2. '6. .-fif-~~* ~ .r. a -
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LICENSE-NO: .DPR - - . A . 2- *
. DOCKET NO:
o 0 - 2 , <,. UNIT NO- l' i
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*i L!CENSEE - E'.*E :. REE LRT -: 0. 9.;-01'? O ?" _
--- 7... .;. -..,..<
nis' Licensee E*zent'Feport is being submitted pursuant _to the-_
-require:.ents_cf:Ccde of Federal Regulation 10CFR50. ~2 ;a r (2 '; i )'(E) .
Issuat. e Of tr.;..>' report is; required
- within_ thirty 301; days of e'/ent-discovery.
Sincerely, n! . A. Ha an Gener l'danager - Salem-operations-MJPJ:vs
- SORC '.94-0 94
;C Distribution LER File
( l nn-s- ww.a . 3,jjg / .
- !~ ..
D
u -- L i f . . . <j_ NRC FORM 366 U.S. NUCLF.AR REGULATORY C0:.iMISSIO$ is s APPROVED BY OMS NO. 315o 0104 , EXPtRES 5/31/95 'I [ ,,-1,, w l astmaarvo susos= paa uts.o si ,o co,u,g, eo.,,,
. UCENSEE EVENT REPORT (LER) - - ro~ cc.u cso , ~ ~ m ,see
- es mouo-2sumoi~ m m to m i e n, ano.-
ano mooses naamaasute emmen a, se ma . v s w:uu segmA.ATo8Pr ookaW=91Mtse mannmeGioN. DC soltsecos. Mb tr } ms Paesawo* wovenon ramtet oiso.oins omcc o. (See reverse for requireo nyrnoer of chg tuenaracters for each D'ock) u**0E** *0 *JJort. w'P**C'oN oc roton b eaceti asus m oocart avusta m en 47 Salem Genetating Station - Unit.1 05000 272 10P 5. tira ia, inacequate Margin tor Pressurizer Overpressure Protection During Low Temperature. Conditions (Applicable to Unit 1)
-l EVENT DATE (t) - i LER NUMBER (6 REPORT NUMBER (7) OTHER FACluTIES INVOLVED (s)
.= i saciurv neat ooc e wuuun uosta ca v. von von se,ma "O*' 0 '""
05000-11 17 94 94 ..-017 - 00 14 94 l l12 05000 OPERAldNG g_ [_THIS REPORT t& SUOMITTED PURSUENT TO THE REQUIREMENTS OF to CFR S: (Cnec= one or more) (11) MODE (9) 20 402ib) 20 405tc) 50 73(a)(2Hiv) i 73 7 t toi POWER 100**
#***""" 50 30'*"" 80#3l'"#"" 73 *"*) '~
j LEVEL 00) 20 405 tam 1HiI 50 36(cH2) 50 73(aH2Hvai x OTMER 20 405<aHilbul 50 73taH?>pi 50 73(a)(2HvmH At 16* " S * * *** :' i 20 405(al(1)bvl . )( 50 7)(ad2Ho) 50 73(a)(2}ivie HB> ", j I 0'. 4 05 'a H 1 h. I , $ } 73lah2Gm) l 50 73(aH2He i 10CFR21
'lCEt4SEE CONT ACT FOR THIS LER (12i n a aet .e v. . ,:.. . -.. co..,
Michael J . Pas t va . J r.- LER Cocrdinator 609 339-5165
~
COMPLETE ONE LINE FOn EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) Oa At s*3 lu Cou*0stA* uass sa:t sa ga ~] f. CassE Systf u Couso% w s usw A;t,r.t o "f "[,8ff f 1
! SUPPLEMENTAL REPORT EXPECTED #146 i EXPECTED I "W" ; ! ' E
*ts . '.- i SUBMISSION -
I h v ves so-we aatPEr .eu ss os oce f DATE (15) '
' ABSTRACT r.mtts yases:es.e arre.~a , , 5 se; e ;3:e: tge A*,r.en timesi 06i On ll/17/94, it was determined that the following realistic assumptions could place Unit 1 outside the design / licensing basis for Pressurizer Overpressure Protection System (POPS) analysis should a_ safety injection (SI) signal occur: Reactor Coolant System (RCS) temperature < 2 0 0*F, 1 l0 Reactor Coolant Pump (RCP) in operation, Positive Displacement Charging Pump (PDP) in service, and power available to-a maximum of 1 Centrifugal Charging-Pump (CCP). Under these conditions, an SI signal could result in a peak RCS pressure of 474 psig from combined flow from the PDP and the CCP, which exceeds the currtnt design basis pressure limit of 450 psig.
This concern-is not applicable to Unit 2. This event occurred becaut-e PDP operation-was not considered in any design basis analysis, based on a POPS lift setting of </= 375 psig. Current operating procedures limit RCP operation in Mode 5 to 1 pump, and require power be removed from the SI pumps in Mode 4 (350*F >Tave, > 2 00'F) . Residual Heat Removal relief valve RH3 is required to be available when RCS temperature is </= 200*F and when the PDP is operating. Code case relief will be requested te increase the pressure / temperature limits operating margin for the POPS ouring Low Tempera are Overpressure conditions. This LER is also intended to satisfy reporting requirements of 10CFR21. seAC sonu 3a4 6 32 . 4 4
- f. .-
' LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Stralon Docket Number LER Number . Pap,e Unit 1 - _
50 272 94-017 2 of 5 Plant and System Identification: Westinghouse - Pressurized Water Reactor Energy Industry Identification ' System (EIIS) codes are shown in the text as {xx} Iderftification of Occurrence 4 Inadequate Margin For Pressurizer overpcessure Protection During-Low Temperature 20nditi:ns .Iq.p;i:21e Tc 'Jnit 1) Event Date: Noved er '.7, 1994 Report Date: :e:e.-bo r .4, 133'
- This report was initiated by ;r.:ider.: Repcrt 94-419 1k This report is intended to also satisfy reporting requirements of 10CFR21.
Initial Conditions: Mode 1 Resctor E:wer ; ., 0 . Ura t Load l'.5; MWe Description of occurrence: The current bases for Technical Specifications (TSs) 3/4.4.9.3 states.that one-Pressurizer Overpressure Protection System (POPS) relief valve, at a lift setting of </= 375 psig, provides adequate relieving capacity in the event of an overpressure transient that includes inadvertent start of a safety injection (SI) pump (mass addition transient) into a water solid Reactor Coolant System (RCS). Subsequently, it has been determined that the following realistic mass addition transient assumptions could place Unit'1 outside the design and licensing basis POPS analysis
- should an SI signal occur:
- RCS temperature </= 200*F
- One Reactor Coolant Pump (RCP) in operation
- Positive Displacement Charging Pump (PDP) in service
- Power supply available to a maximum of One Centrifugal Charging Pump (CCP)
4P t: LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Ds iet Number LER Number Page
- Unit 1 50 272 94-017- 3 of 5 Description of occurrence: -(cont id)
At 1146' hours on-November-17, 199'4- the Ntclear Regulatory Commission was notified.of this event, pursuant to the requirements of 10CFR50.72 (b) (1) (ii) . Under the above condit' ions, an SI signal could result inca combined flow from the PDP and the CCP with a peak =RCS pressure of-474-psig. This exceedsfthe current design bar.is-pressure limit of 450 psig for Salem Unit 1. Analysis of. occurrence; Sac,: ground ' PGPS protects the RCS free exceeding the TS pressure / temperature LP/T) . limits for. plant heatup_! reference 75 Figure _3.4-21 and p: 1 cooldown (reference TS Figure 3.4-3' by opening the Power Operated Relief Valves IP0F")'during low temperature overpressure j i LTOP) transients (RCS cold leg temperamare below 312'F) . Per !' existing design bases, either'of the two PORVs has adequate relieving capacity to protect the BCS.from overpressurl:ation ; when the transient is. limited.to either-(1.) the start _cf an ;dle RCP-with the secondary waterntenperature 'ess than er equal _to 50*F above . the RCS . cold leg . temperature iheat additicni, or - C ' the start of an SI pump and'resultan: _in3ecticn :nto a water solid RCS (mass addition). . ine pressure _ limit at the low temperature end of the P/T curves is presently 450 psig-for Unit 1,.as *ead from the current heatup and cocidown curves. The _ Nuclear Steam Supply System (NSSS) vendor identified in a letter, dated March 15, 1993,.a non-conservatism in the calculation for peak pressure for the-heat input and mass addition transients that affects'both Salem Units 1 and 2. The concern was that the difference between the wide ranr pressure transmitters (PT403 and PT405) elevations, which sen: . hot _ leg pressure and the reactor vessel midplane - (where the TS heatup and
- cooldown P/T limits are defined) with the RCPs operating was not considered in the original Salem POPS analysis. Calculations were performed to address the identified concern and-necessary limits were-met.
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e. LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 4 of 5 Analysis ~of Occurrence: (cont' d) Present Situation Since the satisfactory-completion of this evaluation, it has been determined that the PDP, if already in operation, would continue to operate upon initiation of a SI signal if offsite power-remained:available. During this' postulated event, letdown _would automatically isolate as part-of.the SI actuation. The additional flow from the-PDP is a concern for the limited period of time when the RCS is </= 200*F (Mode 5), the PDP is in operation, and one 11) CCP has its associated power supply available. The ccmbinea flow of 665 gpn fr: -the fDP fl0E gpm) and the CCP (560 gpm, is new c:nsidered the most 1 =: ting mass-addition transient. PSE&G has re-analyzed this dass addition event uring the GOTHIC computer code assuming a bounding maximum pump fitw rate of 675 gpm. The resulting peak pressure is 474 psig, which_exceedsfthe current limit of 450 psig for Salem Unit 1. For the heat addition transient (i.e., the start of an idle:RCP l-with the secondary water temperature </= 50*F above the' RCS cold leg temperature), the peak pressure is 449 psig,-below the PGPS limit of 450 psig.for Salem Unit 1.
-Additional margin on'the TS P/T curves can te :btained when operating-with POPS (RCS cold legs </=.312*F: by taking credit for ASME Code Case N-514. The code case' allows exceeding the P/T limits calculated in accordance with 10CFR50, Appendix G, by 101.
As compensatory action, administrative controls ensure that Residual Heat Removal (RHR) felief valve RH3 is available and the associated RHR isolation valves are in the open_ position. PSE&G has determined that RH3 has similar relieving capacity to that of a PORV. This action is only necessary when RCS temperature is
</= 2 00*F, the PDP is in operation and a CCP has power available.
Apparent Cause of Occurrence: e This event is attributed to " Design", as classified in Appendix B of NUREG-1022. This occurred because the NSSS vendor had not considered the PDP operation as part of any design basis analysis. e O
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- LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
! Salem Generating Station- Docket Number LER Number Page Unit 1 50 272 94-017 5 of 5 Prior Similar Occurrence:
ho other prior similar occurrences have been identified related to this design deficiency. Safety' Significance: Thit event is reportable-in accardance with the requirements of 10CFR50. 73 (a) ( 2 ) (11) (P) , duh to-the POPS not being able to meet its current design bas:s. This event had minimal-safety- 1 signif:.cance, based up;r the sdai:10nal relieving capacity of RH3 ^ an-i/or with the 1 a..: war.'s, p rmitted b'/ use of Code Case . N-514. 1
- Corrective Action:
As interim corrective 3:::n :: ensure compliance with the-' POPS design besis, the foll:w;ng administrative controls are in place:
-Salam Unit 1 operating procedures '.imit RCP operation in Mode 5 to one pump, and require-that power be . removed from - the SI pumps l in Mode 4- (35:*r >Tave, . r: : T . VE3 is required to be available i to ensure the :urrer.: ?'T .;-its ar's met when RCS temperature is
$/ = 2 00 F, when the P;l :s .r Operation. Similar administrative controls for Sale 'init 2 are in place, although RH3 is.not-needed'to meet the curresponding Unit 2 P/T limit.
l -- A1submittal will be made to the NRC requesting. application of . Code Case N-514. This additional margin increases the P/T limits operating margin for the PCPS during LTOP conditions. fi J. J. n General Manager Salem Operations MJPJ REF: SORC Mtg. 94-094 l m ;
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REACTOR COOLANT SYSTEM mu rs -
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Finally. ; the _new 10CFR50 rule which addresses the metal temperature: of the closure head > flange regions is_ considered. This 10CTR50 rule' states that the-metal- temperatuge- of- the closurei flange' regions aust exceed- the material-RT4 byat.least:120Fifor-normaloperationwhen:thepressure_ exceeds 20 percent'! !- the.preservicehydrostatic-testpressure(621pagg-forSales). . Table-
.53/4.4 1 indicates that the limitt,ng-RT . of 26 F occurs in-the closure he'ad is- 148'F at pressures greater than 621.psig. - These-limits do' not aff s
Figures 3;4 2 and__3.4 3. 1 Although the. pressurizer operates in temperature "rangee above those for awhich - there istreason for concern of non ductile! failure.l operating limits _are
-provided_to assure s compatibility.of operation with:the fatigue analysis
-performed in_accordance with the-ASME-Code requirements; The OPERABILITY of two POPSs or-an RCS vent cpening of greater-than.3.14:
-square inches.. ensures that the RCS_will be protectd from pressure transients ,
which could exceed che . limits of Appendix C -co 10 gm Part 50-when one or more. of the RCS_ cold legs are less tha'n or equal to 312-.F._ EitherfPOPS has= { adequate relieving capability to protect the RCS from overpressurization-when thel transient = is- limited to' either (1) - the start of an-idle.- RCP ~with the secondary water cesperature of the : steam' generator less - than:or= equal to 50*T ;
- above the RCS cold 11eg' temperatures.-or_(2) the start of
~
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4 A testing provision in the POPS circuitry allows for test opening of the relief valves prior to use of the systern below 3120F, The
" TEST
- pushbutton, when depressed, will operate the relief valve provided that the associated motor operated valve is closed.
Other portions of the POPS can be tested in a manner similar
- to a 6 #
other protection system functions. The existing power operated relief valves (PCRVs) are utilized for overpressure , protection at low temperature in Units 1 and 2. 7.6.3.3 Desien Evaluation The POPS is designed as a " protection grade" system in accordance with the applicable portions of IEEE Standard 279 1971. The use of proven devices provides assurance that the system is compatible with other Protection System equipment. The use of ad:ninistrative controls to arm the POPS is considered acceptable due to the expected infrequent need for overpressure protection at low , temperature. , The POPS relief valves protect the RCS from pressure transients l ! wnich could exceed the limits of Appendix C to 10CFR50 when one or l more RCS cold leg temperature is at or below 3120F. Either POPS has 3 adequate relieving capacity to protect the RCS from overpressurization as a result of the limiting heat input or mass input cases: (1) the start of an idle Reactor Coolar.t Pump with-the secondary water temperature of the steam generator less than g;f or equal to 500F above RCS cold leg temperature or (2) the start of g safety injection pump and its injection into the water solid RCS
'-Y number of provisions for prevention of pressure transients below P-7 (when the RCS temperature is below 3120F) presently exists in the Technical Specifications.
In order to cause an unwanted relief valve opening at normal operating pressures, an operator would have to erroneously arm the POPS system. This would require bypassing the administrative I control of the key associated with the keylocked pushbutton A Qt Gt 2 SCS-UFSAR (W,{f. A ion 6
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1034-U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn.: Document Control Desk SALEM GENERATING STATION .
. LICENSE NO: DPR-70 DOCKET NO: 50-272
= UNIT NO: 1 of LICENSEE EVENT REPORT NO. 94-017- M This Licensee Event Report is being subritted pursuant .to the requirements of Code of Federal Regulation 10CFR50.73 (a) (2 ) (4 (B). Incuance n# thic repcrt --'" req"i"ad c
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1 i LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page Unikt 4 50-272 94-017-6 l 2 of 5 Plant and System Identification: Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are shown in the text 'is (xx) Identification of Occurrence: Inadequate Margin For Pressurizer. Overpressure Protection During Low Temperature Conditions (Applicable To Units 1+ 2. , 7 Event Date: November 17, 1994 A ki35,- Report Date: %ccMer D l', ~ This report was initiated by Incident Report 94-419 This rep st is intended to a'en wi rfy reporting r quiremente of % 10^FR21. ' ~ Initial Conditions: ( Mode 1 Reactor Power 1001, Unit Load 1150 MWe s [d' Description-of Occurrence: fg The current bases for. Technical Specifications (TSs) 3/4.4.9.3(r r,- , [ states that one Pressurizer Overpressure Protection-System (POPS){ ]
' relief-valve, at a lift setting of </= 375 psig, provides adequate-relieving capacity in the event of an overpressure transient that includes inadvertent start of a safety injection (SI) pump (mass addition transient) into a water solid Reactor Coolant System (RCS). Subsequently, it has been determined that
. -the following realistic mass addition transient assumptions could plac7 Urn 4 l. outside the design and licensing basis POPS analysis should an SI signal occur:
- RCS temperature </= 200*F
- One Reactor Coolant Pump (RCP) in operation
- Positive Displacement Charging Pump (PDP) in service
- Power supply available to a maximum of One Centrifugal Charcing Pump. (CCP)
f C ? e: I- . 1 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 tol 3 of 5 Description of Occurrence: (cont' d) At'1746 hours on November 17, 1994, the Nuclear Regulatory Commission was notified of this event, pursuant to the requirements of 10CFR50.72 (b) (1) (ii) . Under the above conditions, an SI signal could result in a combined flow from the.PDP and the CCP with a peak RCS pressure of 474 psig; This exceeds the current design basis pressure limit of-450 psig for Salem Unit 1.
^M V g M M 2_ gq q Analysis of Occurrence:
Background
POPS protects the RCS from exceeding the TS pressure / temperature (P/T) limits for plant heatup (reference TS Figure 3.4-2) and cooldown (reference TS Figure 3.4-3) by opening the Power Operated Relief Valves (PORV) during low temperature overpressure (LTOP) transients (RCS cold leg temperature below 312*F) . Per existing design bases, either of the two PORVs has adequate relieving capacity to protect the RCS from overpressurization when the transient is limited to either- (1) ,the start of an idle RCP with the secondary water temperature less than or equal to-50'E above the RCS cold leg temperature (heat addition), or (2) the start of an SI pump and resultant-injection into a water solid RCS (mass addition). The pressure limit at the low temperature end of the P/T curves is presently 450 psig for Unit as 1, Ac read from the current heatup and '.% / li on5"M M l1 # 2_ e pcooldown
,..x + q pcurves.
( .1,, v41. .C %: . w. The Nuclear Steam Supply System (NSSS) vendor identified in a A,.... . letter, dated March 15, 1993, a non-conservatism in the +. .- .R calculation for peak pressure for the heat input and mass u.w ... addition transients that affects both Salem Units 1 and 2. The concern was that the difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg , pressure and the reactor vessel midplane (where the TS heatup and cooldown P/T limits are defined) with the RCPs operating was not considered in the original Salem POPS analysis. Calculations were performed to address the identified concern and necessary limits were met.
g. g.: LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number -Page Unit 1 50-272 94-017 4 of 5 Ana%ysis of Occurrence: (cont' d) Present Situation Since the satisfactory completion of this evaluation, it has been determined that the PDP,- if already in operation, would continue to operate upon initiation of a SI signal if offsite power remained available. During this postulated event, letdown wculd automatically isolate as.part of'the SI actuation. The additional flow from the PDP is a concern for the limited period of time when the RCS is </= 200*F (Mode 5), the PDP is in operation, and one (1) CCP har its associated power supply
~available. The combined flow of 665 gpm from the PDP (105 gpm)
I and the CCP (560 gpm) is now considered the most limiting mass ! : addition transient. PSE&G.has re-analyzed this mass addition event using the GOTHIC !. computer code assuming a bounding maximum pump flow rate of 675 gpm. current.ne resulting limit of 450 peak psigpressure is 474 for Salem Unitpsig
- 1. @, which exceeds the PQQ f
.For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature </= 50*F above the RCS cold below the POPS leg. temperature), the peak pressure is 449 psig d"t d I W M .b Ei limit;5 o f '50 pcir for Sal * * * ~. M ."Ht 1. M 2. g M. .g
' Additional margin on the TS P/T curves can be obtained when operating with POPS (RCS cold legs </= 312*F) by taking credit _
for ASME Code Case N-514. The code case allows exceeding the P/T limits calculated in accordance with 10CFR50, Appendix G, by lot. As compensatory action, administrative controls ensure that Residual Heat Removal (RHR) relief valve RH3 is available and the associated RHR isolation valves are in the open position. PSE&G has determined that RH3 has similar relieving capacity to that of a PORV. This action is only necessary when RCS temperature is
</= 200 F, the PDP is in operation and a CCP has power available. 3 (A IO CEA$o.fi @, u ' - A % W W/M k @ @-tlu m5{
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Apparent Cause of Occurrence: This event is attributed to " Design", as classified in Appendix B of NUREG-102?. This occurred because the NSSS vendor had not considered the PDP operation as part of any design basis analysis.
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7 U' 1. LICENSEE EVENT REPORT (LER) TEXT CONTINUATION-Salem Generating Station - Docket Number ~ LER Number Page Unit 1 50-272 94-01') 5 or5 a Prior Similar Occurrence: No other prior similar occurrences have been identified related to this design deficiency. Safety Eignificance: , This event is reportable in accordance with the, requirements-of F' * ,, 10CFR50. ~13 (a) (2) (ii) (B) , due,tothePOPSnot.beingabletomeept ~ lts-current design basis. This event had minimal safety ,o y significance, based upon the additional relieving capacity of RH3-
~
and/or with the 10% allowance, permitted by use of Code Case
,N-514.
t: Corrective Action: Xs interim corrective action to ensure compliance with the POPS design-basis, the following administrative controls are in place: l Salem Unit 1 operating' procedures limit RCP operation in-Mode 5 to one pump, and require that power be removed from the SI pumps in Mode 4 (350'r >Tave, > 2 00'r) . RH3 is required to be available to ensure the current P/T limits are met when RCS temperature is
</= 200*F, when the PDP is in operation; Similar administrative j controls for' Salem Unit 2 are in place, although RH3 is not I needed to meet the corresponding Unit 2 P/T limit.
A submittal will be made to the NRC-requesting application of b Code Case N-514. This additional margin increases the P/T limits operating margin for the POPS during LTOP conditions. W SO .pt p - 'T, r t
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J. J. an General Manager Salem Operations MJPJ REF: SORC Mtg. 94-094
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-U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn.: Document Control Desk SALEM GENERATING STATION LICENSE NO: DPR-70 DOCKET NO: 50-272 UNIT NO: 1 SUPPLEMENTAL LICENSEE EVENT REPCRT NO. 94-017-01 l
This Licensee Event Report supplement is being submitted pursuan , to requirements of the Code of Federal Regulations 10CFR50.73. It addresses the applicability of the original issue to Unit 2 and also provides additional information related to root cause and corrective actions. Sincerel , Clay C. Warren General Manager - - Salem Operations MJPJ:vs SCRC 95-098 C Distribution
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rasiuTT - 49 vessui- :::nin enes is SALEM GENERATING STATION UNIT 1 06000272 1 of 7 TITb5 to) INADEQUATE MARGIN TEMPERATURE CONDITIONS FOR PRESSURIZER OVERPRESSURE PROTECTION DURING LOW EVENT DATE (3) LER NutleER (5) REPORT OATE (7) oTHER F ACILITIE 5 INVOLVEb (8) y esowtM Dat t:4A #
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~ ~ F acauTl femals DoGRET eeWelHR 1 1 17 94 94 017 0 1 08 2 5 95 OPERATING 05000 THIS REPORT I5 SuettlTTED teo0E(9) 24.2201(b) 'UR5UhWT TO THE fl20UIREteENTS OF 10 CFR lt (Check one or merel (11) 20.2203(e)(2Hv) 50.73(e H 2)(i) 50.73(o)(2)(vne)
-@wiR 20.2203(e)(1) 20.2203(eH 3HO X 80.73(e H 2Hii) 50.73(e H2Wal LEVEL (10) 20.2203(eH2)(Q 20.2203(e H3)(u) 50.73(e H 2)(na) 73.71
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LICE NSI.E CONTACT FOR THis LER (12) aus r arnoas awasesr ca. ar c.e.i CRAlOLAM8ERT 609/339 1848 l-l i COMPLETE OME UNE FOR EACH COEONENT FAILURE 045CRieEO IN THIS REPORT (13) caves erstem compo*4we mauveaervaan ago,a,tg g cavu sistem commo=sur namweactuate ago,y,*,e,a
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SUPPLEa8 ENTAL REPORT EXPECTED (14) EXPECTED mon 1H DAY VEAR YES NO SUBMtSSCH tw yes, complete EXPECTED SueWIS$loN DATE). X DATE (15) Ae5 TRACT tLima to 1400 epeces.i.e . ppro meieiy'18 emei.. pec a typewraten smn) (1st On 11/17/94, a 10CFR50.72 (b) (1) (ii) notification was made when it was determined that based on realistic assumptions far various sys. tam , availability and alignments, Salem Generating Station Unit 1 was outside the design / licensing basis for Pressurizer overpressure Protection System (POPS) should an inadvertent safety injection (SI) . signal occur in Mode 5 below 200* F. Under these conditions, an SI signal could result in a peak RCS pressure of 474 psig which exceeded the then current design basis pressure limit of 450 psig. This concern was not considered applicable to Unit 2 at that time. Fubsequent review of an engineering evaluation has determined that both Units 1 and 2 were outside their design / licensing basis for the POPS analysis and should have been reported on December 30, 1993. This conclusion is based on the differential pressure from the-cperating RCPs that was not considered in the original POPS analysis for the mass addition transient. The transient analysis code and inappropriate use of ASME Code Case N-514 led to the misreporting. Code Case N-514 has cince been approved by the NRC for use at Salem. wac ronM 3es e est r
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"ERATING STATION UNIT 1 05000272 2 OF 7 94 - 017 - 01
,nero ep co.e reewees. wee .aow.ai copies of NRC f orm 366A) (17) idnt and system Identificati 2A1 W)stinghouse-PressurizedWaterReactor
,LEnergy Ind .try Identification System (EIIS) codes are shown in the text as
,., (xx) !
1 Identification of occurrence: Inadequate Margin For Pressurizer overpressbre Plotectien During Low Temperature conditions (Applicable To Units 1 and 2) Event Dates November 17, 1994 and December 30, 1993 Initial Report Date December 14, 1994 Report Date August 25, 1995 This report was initiated by Incident Reports 94-419 and 95-343. Initial conditions: Mode 1 Reactor Power 100% Unit Load 1150 MWe Descrintion of occurrenc.91 The hases for Salem Units 1 and 2 Technical Specifications (TSs) state that one Pressurizer overpressure Protection System (POPS) relief valve, at a lift setting of </= 375 psig, provides adequate relieving capacity in the event of an overpressure transient that includes inadvertent start of a safety injection (SI) pump (mass addition transient) into A water solid Reactor Coolant System (RCS). Subsequently, it was determined that the following realistic mass addition transient assumptions could place Unit 1 outside the design and licensing basis POPS analysis should an SI signal occur
- RCS temperature </= 200' F
- One reactor coolant pump (RCP) in operation
- Positive displacement Charging Pump (PDP) in service
- Power available to a maximum of one Cantrifugal Charging Pump (CCP)
At 1746 hours on November 17, 1994, the Nuclear Regulatory Commission was notified of thit, event relative to Unit 1, pursuant to the requirements of 10CFR50. 7 2 (b) (1) (ii) . NRC FOMM 3e64 (606) ,
t u.s. wuota.AA accutAtony commisseou 4 M s
. UCENSEE EVENT REPORT (LER)
', A TEXT CSNTINUATION me i
- # _Len eeutsetn to) . . . Pact (a)
N ERATING STATION UNIT 1 a **11.7 L*'*; 05000272 94 - 017 - 01 3 OF 7
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.,.4, u.. .ao. con., cop... omac rarm mA) p F) qcurrencer icont'dl n$artheaboveconditions,
,J from the PDP and CCP with a peak RCS pressure of 474 psig.an This exceeded SI signal cou 4
the current design basis pressure limit of 450 psig for Salem Unit 1. Subsequently, review of an engineering evaluation completed on December 30, RCPs Units would 1 and 2 have resulted in the pressure / temperature limits for both (P/T)1993, j
- 5. (450 psig and 475 psig respectively) being exceeded in Mode This determination is based on the differential precsure from the RCPs not being considered in the original POPS analysis for the mass addition transient. i I
Analysis of_occurronen.t POPS (reference protects the RCS TS Figure 3.4-2)from exccedir.g the TS P/T limits for plant heatup opening the Power Operated Relief Valvesand cooldown (reference TS Figure 3.4-3) by overpressure (LTOP) transients (PORV) during low tnaperature Per existing design bases, (RCS cold leg temperature below 312' F) . capacity to protect the RCS from overpressurization when the transient iseith limited to either (1) the start of an idle RCP with the secondary water temperature less than or equal to 50' F above the RCS cold leg temperature (heat into addition), a water solid or RCS (2) the start of an SI pump and resultant injection (mass addition . The pressure limit at the low temperature Unit 2, end of the P/T curves was)450 psig for Unit 1 and 475 psig for 30, 1993.as read from the heatup and co71down curves in effect on December The original Sslem POPS analysis, based on the LOPTRAN computer code, calculated a maximum peak pressure for the most limiting mass addition transient analysis, of 446 psig with the PORV set at a pressure of 375 psig. In this initially cold water rolid RCS was considered.the RCS pressure due to injection of 7 The Nuclear Mcrch 15, 1993,Steam Supply System (NSSS) vendor identified in a letter, dated a non-conservatism in the calculation for peak pressure for the 2. ond heat input and mass addition transients that affects both Salem Units 1 transmitters (PT403 and PT405)The concern was that the difference between the vide r elevations, which sense hot leg pressure, and the reactor limits are defined) vessel midplane-(where the TS heatup and cooldovn P/T original Salem POPS analysis,with the RCPs operating was not considered in the i
py g .,
, u.* aucuAn natGutarony commessaoa y
g , CENSEE EVENT REPORT (LER) TEXT CONTINUATION oo_cajT_aguaitm_122 f- - ma un numeER L mL TING STATION UNIT 1
, ".E? TC:sf:'
05000272 4
' 94 - 017 - 01 OF 7 f- e,.
.i , guir.e, ves soa cono copies et Nac rorm usa) o r) 17 sis of Occurrences (cont'd)
' quantify
, operation were thedetermined effects, specific for one, pressure differences associated with RCP <
two, and four The 77 results one, of these calculations provided values of 29,RCPs 37, operating. and 71 psig with 4 two, and four RCPs operating, respectively. 2.0 psig was then added to account for transmitter elevation differencesA correction pressure of not previously accounted for in the original calculations. When considering the pressure differential from one RCP in operation (31 psig) and adding this pressure to the peak pressurn for the mass addition exceeded. (446 transient psig), the P/T limits for both Salem Units 1 cnd 2 were Therefore, reporting purposes. both units were outside of their design basis for It was also determined that the PDP, if already in operation, would continue remained available. to operate upon initiation of a SI signal if offsite power During this postulated event, letdown would l outomatically isolate as part of the SI actuation. The additional flow from the PDP is a concern for the limited period of time when the RCS is
</= 200* F (Mode 5) ,
ossociated power supply available.the PDP is in operation, and one (1) CCP has its PDP (105 gpm) and the CCP (560 gpm) The combined flow of 665 gpm from the nass addition transient. is now considered the most limiting PSE&G has re-analyzed this masa addition event using the GOTHIC computer code assuming a bounding maxiuum pump flow rate of 675 gpm. peak Unit 1, pressure is 474 psig, which exceeded the limit of 450 psig for SalemThe resulting but was within the limit of 475 psig for Unit 2. Additional margin on the TS P/T curves can be obtained when operating with POPS (RCS cold legs </- 312' F) by applying ASME Code Case N-514. The code case allows exceeding the P/T limits calculated in accordance with 10CFR50, Appendix G, by 10%. PSE&G requested permission to utilize Code Case N-514 and the NRC approved the request on 2/13/95. On December 22, 1994, a 10CFR50.59 Safety Evaluation vac completed for Salem Unit 2 that changed the POPS TS Bases. The mass addition flow rate ecsumed for the present POPS analysis is limited to the combined flow from f the CCP in conjunction with an operating PDP or SI pump while in Mode 5. On February 8, 1995, a 10CFR50.5) Safety Evaluation was completed for Salem Unit 1 that changed the POPS Technical Specification Bases. The mass addition single CCP flow while assumedin Mode for5.the POPS analysi's was-limited to the flow from a by the NRC, Following approval of the ASME Code Case N-514 changed the Technical Specification Bases.an additional 50.59 Safety Evaluation was complete The new mass addition flow e m Q
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==1vais of Occurrence (cont'd):
A rate assumed for the POPS analysis is limited to the combined flow from the , CCP in conjunction with an operating PDP or SI pump while in Mode 5. This is identical to Uni
- 2 Bases. i
'l Accarent cause of Occuryoncet This event is attributed to Design, as classified in Appendix B of NUREG-1022. This occurred because the NSSS vendor had not considered either the pressure differential associated with the operation of the RCPs or PDP operation as part of the design basis analysis for the mass addition transient.
In December 1993, PSE&G inappropriately utilized the 10% margin allowed by code Case N-514 without prior NRC approval. Ucilizing the additional margin allowed by the code Case resulted in failure to recognize that the P/T limits could be exceeded for both Units. PSE&G also failed to in'itiate timely or effective corrective action in accordance with Nuclear Administrative Procedure NC.NA-AA.ZZ-0006(Q) to address the POPS non-conservatism after being notified by the NSSS vendor in March 1993. Also, PSE&G credited analysis results for the mass addition transient utilizing the GOTHIC computer code rather than LOFTRAN (used for the original mass cddition transient) without the completion of a 10CFR50.59 safety ovaluation. All of these factors resulted in PSE&G's failure to recognize that both Salem Units were outside their design bases in December 1993, , when considering the pressure differential associated with one or more RCPs in operation. Prior Similar Occurrencet No other prior similar occurrences have been identified related to this design deficiency. 3.itity_81onificance This event is reportable in accordance with the requirements of 10CFR50. 73 (a) (2) (ii) (B) , due to the POPS not being able to meet its current design basis. This event had minimal safety significance, based upon the additional relieving capacity available through the use of RH3 and/or with the 10% allowance. permitted by use of Code case N-514. WCAP - 13366, " Analysis of Capsule X From PSE&G Salem Unit 2 Reactor Vessel i Radiation Surveillance Program", dated June 1992, analyzed the effects of radiation on the Unit 2 reactor vess11 to determine the impact on the opt, rating P/T limits. The results of that analysis determined that at an
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Safety Sionificance contid fr NI l
-, RCS temperature of 85* F the pressure is 495 psig. In January 1993. PSEL3 l cubmitted a TS change request to modify the P/T limits based on the WCAP. I The change request was approved by the NRC in February 1994 and made l Offective in April 1994. '
i The change in the pressure limit at the low temperature erd of the P/T l curve, while not approved by the NRC in December 1993, shows that in ' actuality the P/T limits for Unit 2 were never exceeded and the original i concerns identified on Unit 1 were not applicable to Unit 2. Additionally,! ' the revised P/T limits for Unit 2 and the new limits provided by the ' additional 10% ma: gin allowed by Code Case N-514 ensures that the current - l P/T limits for Salem Units 1 and 2 will not be exceeded for an LTOP event, l l i Correceive Actient The fol.cwing administrative controls are in place on Salem Units 1 and 2 to ensure compliance with the PCPS analysis: ,
- 1. Procedures limit cperation in Mode 5 to two (2) RCPs.
- 2. Pcwer must be removed frem the S: pumps upcn entry into i Mode 4 ' 3 5 0' F > Tave > 2 0 0' F) .
A submittal was made to the NRC requesting permission to uti.ize ASME Ccde Case N-514 to allow an addit cnal margin of 10% in the P/T limits for the ' PCFS during LTCP conditiens. The NRC approved PSELG use of Code Case N-514' en February 13, 1995, f The Corrective Action Program has been significantly improved by combining i the previous processes for reporting conditions adverse to quality, lowering the program threshold, formalizing the operability Determinati:n , Precess, increasing management involvement and oversight, and clearly : communicating management expectations regarding timeliness of evaluaticns ! and corrective actions. l 1
'.i Management has re-emphasized supervisions primary role to assess emerging ,
issues objectively, as opposed to helping develop a solution. i Procedure and progra and compliance has been re-emphasized. especially in the a{ps 'cfcmmitment Corrective Actions. Management has re-emphasized that 10CFR50.59 is applicable if revisions te ! calculations / evaluations alter either the design basis, basis of analysis - or conclusions in the UFSAR. l,' I
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i,-" P responsibility for compliance with licensing commitments and p reporting. Guidance ASME Code has been provided application. to appropriate Engineering personnel regarding A 10CFR50.59 safety evaluation vill be completed to allow PSE&G to utilize the GOTHIC completed masscomputer additioncode rather than LOFTRAN for the most recently transient.
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i vill be implemented to ensure OEF documentsProcess (e.g. improvements Program to the into the Corrective Action Program. receive initial screening for operability and I Isr,ves that involve potential operability Program. concerns vill be prioritized through the corrective Action The Technical Specification Bases for Units 1 and 2 have been revised to reflect the chenges in the assumptions for the mass addition transient. controls arechanges Procedure maintained. have also been implemented to assure that appropriate REF: SORC Mtg. 95-098
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