ML20207L683
| ML20207L683 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 03/10/1999 |
| From: | Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9903180224 | |
| Download: ML20207L683 (6) | |
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.PECG NUCLEAR nemoc-<
Nuclear Group Headquarters A UNir or PfCO ENfRGY 965 ChestertmA Boulevard i
I Wayne. PA 19087-5691 i
March 10,1999 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 i
l U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555
' subject: Limerick Generating Station, Units and 1 and 2 Response to Request for Supplemental Information Technical Specifications Change Request No. 98-08-0
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Increase Safety Relief Valve Setpoint Tolerance From 11% 1013%
Dear Sir / Madam:
By letter dated January 12,1999, PECO Energy Company submitted Technical Specifications (TS) Change Request No. 98-08-0, in accedance with 10CFR50.90, requesting an amendment to the TS (Appendix A) for Facility Oper. ' *o License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2. The proposed changes involve revising TS Section 3/4.4.2, " Safety / Relief Valves," and TS Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2, to increase the allowable as-found main steam Safety Relief Valve (SRV) code safety function lift setpoint tolerance from *1% to 13%.
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Subsequently, during a telephone conversation on March 2,1999, the NRC requested 1
clarification regarding a number of issues pertaining to the proposed TS change. During this 1
1 conversation, several of the NRC's issues were resolved. However, the NRC requested that PECO Energy submit supplemental information on five (5) specific issues in order for the NRO to
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continue its review of the proposed TS Change Request.
Accordingly, the attachment to this letter provides PECO Energy's response to the specific issues. Each issue is restated fodowed by our response. This request is being submitted under affirmation, and the required affidavit is enclosed.
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..- n n G 9903190224 990310 PDR ADOCK 05000352
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March 10,1999 Pege 2 if you have any questions, piease do not hesitate to contact us.
Very truly yours,
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arrett D. Edwards Director-Licensing Attachment Enclosure cc:
H. J. Miller, Administrator, Region I, USNRC (w/ enclosure)
A. L. Burritt, USNRC Senior Resident inspector, LGS (w/ enclosure)
R. R. Janati, PA Bureau of Radiological Protection (w/ enclosure)
J COMMONWEALTH OF PENNSYLVANIA
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' COUNTY OF CHESTER 1
J. J. Hagan, being first duly sworn, deposes and says:
That he is Senior Vice President of PECO Energy Company, the Applicant herein; that he has read the foregoing request to provide supplemental information in support of the proposed amendment requests to Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station, Units 1 and 2, concerning Technical Specifications Change Request No. 98-08-0," increase Safety Relief Valve Setpoint Tolerance From 11% 1013%," and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.
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.M r Vic resident Subscribed and sworn to before me this /B day o
1999.
Ah N6tary Public NotarN seg I- - -
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Dock t Nos. 50-352/353 Attachm:nt March 10,1999 Page 1 of 3 Limerick Generating Station, Units 1 and 2 R9sponse to Request for Supplemental Information Regarding Technical Specifications Change Request 98-08-0 l
NRC Question
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In the January 12,1999 submittal on page 5 of 15, first paragraph RCIC System it is stated that 1
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"The RCIC system is designed to reach rated flow within 30 seconds; however the response time i
usedin the analysis is 55 seconds. Therefore, the response time is negligible." The increase from 30 seconds to 55 seconds is a substantial change. Clarify where this assumption is used and what willbe the impact if 30 seconds is usedinstead of 55 seconds. Page 6-4 of Enclosure-1, GE report, firstparagraph-If RCIC is designed to be capable ofinjecting rated flow of 600 gpm within 30 seconds, why the TS limit is i<ept at 55 seconds?
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Response
The Limerick Generating Station (LGS), Units 1 and 2, Updated Final Safety Analysis Report (UFSAR),
l Section 5.4.6, " Reactor Core isolation Cooling," states in part that: The RCIC system is designed to initiate and discharge, within 55 seconds, a specified constant flow into the reactor vessel over a specified
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pressure range.
The RCIC system design specification requires that "the RCIC system shall start automatically and deliver design flow within 30 seconds".
As part of the SAFER /GESTR-LOCA PERFORM analysis, an additional 25 seconds of margin to the start-up requirement was added. This change from 30 seconds to 55 seconds was performed under 10CFR50.59 and implemented in 1995.
i NRC Question (Related to GE Topical Report NEDC-32645P)
Page S-2, Fifth paragraph-ATWS-Usually MSIV closure is the limiting case for ATWS. Why PREGO eventlimiting?
Response
MSIV Closure and Pressure Regulator Failure - Open (PREGO) are similar events, and either event has the potential to result in the maximum calculated overpressure. For this reason, both events were l
considered. The MSIV closure event assumes that all MSIVs close simultaneously while the unit is operating at rated conditions. The PREGO event assumes that the pressure regulator fails open, resulting in a maximum steam demand. This maximum steam demand results in a reduced steam pressure causing the MSIVs to close on iow steamline pressure. Therefore, the MSIV closure during a PREGO event occurs at conditions other than rated.
Previous evaluations using the REDY code identified the MSIV closure event as limiting for the ATWS c /erpressure analysis, as documented in General Electric Report NEDC-32265P, Limerick Generating Sation Units 1 and 2 Power Rerate Engineering Report. The supporting analysis for this submittal used l
the OYDN code for the ATWS overpressure analysis and identified the PREGO event as limiting for l
overpressure, as documented in GE Report NEDC-32645P, Limerick Generating Station Units 1 and 2 SRV Setpoint Tolerance Relaxation Licensing Report. Both codes are approved by the NRC for the ATWS overpressure evaluation.
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Dockit Nos. 50-352/353 Attchment March 10,1999 Page 2 of 3 E
NRC Question (Rotated to GE Topical Report NEDC-32645P)
Page 6-1.1 HPCI SYSTEM-The current performance design basis for the HPCI system is that it must be capable ofinjecting design rated flow into the reactor vessel at a maximum reactor pressure squal to the lowest SRV nominal set point plus the allowable set point tolerance. And in LGS USAR Table 6.3-1A, Page 2 of 3 HPCI minimum t;ow rate independent of vesselpressure is specified as 5600 gpm. Obviously, this requirement is not satisfied with proposed TS change where 5000 gpm is specified for reactor pressure between 1182 psig and 1205 psig. What is the design flow for HPCI now? If only 5000 gpm is required for satisfying 10CFR 50. 46, FSAR sections should be changed to show that a minirnum of 5000 gpm is sufficient.
Response
The HPCI system design flow requirement will be changed following approval of this proposed Technical i
Specification change. The current design flow for the HPCI system is 5600 gpm at reactor pressures between 200 to 1182 psig. The new design requirements for the HPCI System flow as a result of this submittal will be 5600 gpm at reactor pressures between 200 to 1182 psig and 5000 gpm at reactor pressures between 1182 to 1205 psig.
The change in the design flow for the HPCI system is desired to avoid increasing the maximum rated pump speed which results in a condition that may lead to a HPCI system overpressure concem. If the j
current design flow of the 5600 gpm was to be maintained at the increased reactor pressure of 1205 psig, the HPCI turbine / pump speed would need to be increased. As a result of this turbine / pump speed increase, some HPCI system components may exceed American Society of Mechanical Engineers (ASMC Oode allowables under a postulated failure of the injection valve. By maintaining the current turbine / pump speed settings, all components will remain within allowable limits. However, due to the increased reactor pressure from 1182 psig to 1205 psig, the system operating pressure and pump Total Dynamic Head (TDH) is increased resulting in the potential for reduced HPCI system flow as pump operation follows the constant speed pump curve.
Surveillance testing of the HPCI system will continue to be performed in accordance with the Technical Specification Surveillance Requirements.
The HPCI system flow input parameter used in the current LGS, Units 1 and 2, SAFER /GESTR-LOCA Analysis (i.e., GE Report NEDC-32170P," SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis for Limerick Generating Station Units 1 and 2"), which demonstrates compliance with the requirements of 10CFR50.46, stipulates a HPCI system flow rate of 5400 gpm at reactor pressures between 200 to 1141 psig. Both the current HPCI system design flow of 5600 gpm at reactor pressures between 200 to 1182 psig, and the proposed HPCI system design flow of 5600 gpm at reactor pressures between 200 to 1182 psig and 5000 gpm at reactor pressures between 1182 to 1205 psig meet the requirements of 10CFR50.46.
For events that result in reactor pre asures greater than 1182 psig a HPCI system flow of 5000 gpm has been determined to be adequate. kose events include, but are not limited to, plant transients and special events such as Station Blackout, Fire Safe Shutdown, and High Energy Line Breaks. Loss-of-Coolant Accidents (LOCAs) are expected to result in reactor pressures less titan 1141 psig.
The LGS, Units 1 and 2. UFSAR will be updated to reflect these revised HPCI flow rate values, following issuance of the amendments.
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Docket Nos. 50-352/353 Att: chm:nt March 10,1999 Page 3 of 3 NRC Question (Related to GE Topical Report NEDC-32645P)
Page 6-2, third para-Clarification is required for the statement that '-relaxed flow rates for the HPCI System are consistent with the current ECCS-LOCA analysis."Now HPCI needs to be operated at reactorpressure of 1205 psig. The flow is rvducec' to 5000 gpm, hence the current ECCS-LOCA analysisis changed.
Response
l The HPCI system flow input parameter used in the current LGS, Units 1 and 2, SAFER /GESTR-LOCA Analysis (i.e., GE Report NEDC-32170P, SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis for Limerick Generating Station Units 1 and 2), which demonstrates compliance with the requirements of 10CFR50.46, stipulates a HPCI system flow rate of 5400 gpm at reactor pressures between 200 to 1141 psig. Both the current HPCI system design flow of 5600 gpm at reactor pressures between 200 to 1182 psig, and the proposed HPCI system design flow of 5600 gpm at reactor pressures between 200 to 1182 psig and 5000 gpm at reactor pressures between 1182 to 1205 psig satisfy the requirements of 10CFR50.46. Therefore, the current analysis is not changed.
NRC Question Page 7 of Attachment 1 refers to a manufacturing uncertainty of 5% to calculate the SRV steam flow, and this appears to be treated as additional margin to counteract the additional 3% to the setpoint pressure. However, since the consideration of a manufacturing uncertainty implies that the additional 5% to the SRV flow can actually exist, and it would therefore be inappropriate to remove it to address the effects of the additional 3% to the setpoint pressure. Is there additional structuralmargin (e.g., due to overpredicting the forces resulting from the SRV flows) which could be credited?
Response
The standard method used to calculate steam flow though a SRV at various reactor pressures uses nominal SRV setpoints. This method is presented in GE Report NEDO-21888, Mark l Containment Load Definition Report. The method specified in NEDO-21888 was used for developing the SRV load definition. NEDO-21888 states the following:
"The SRV flow rate is assumed to be 1.225 times the ASME rated S/RV flow. This factor was obtained as follows: A factor of (1/0.9 )is applied to the ASME rated flow rate to account for the fact that tilis value is by definition only 90 percent of the expected S/RV flow. In addition, a factor of 1.05 is applied to account for the allowable uncertataty of the S/RV loss coefficient, K p. Finally, a factor of 1.05 is applied for conservatism. The 1.225 multiplier is the product of the factors described above, i.e., (1/0.9) X 1.05 X 1.05 = 1.225."
Since the current nominal SRV setpoint values are not being changed and the revised setpoint tolerance is less than 5%, the calculated SRV steam flow was not impacted. Therefore, this proposed change will have no impact on the SRV hydrodynamic loads on the suppression pool wc:1 or submerged structures.
Additionally, the LGS SRV quencher loads are based on actual test data and generic quencher loads based on a reactor vessel pressure of 1276 psig. Applying the maximum setpoint tolerance of +3% to the highest nominal SRV setpoint of 1190 psig yields a maximum reactor vessel pressure of 1226 psig which is bounded by the design reactor vessel pressure for quencher loads of 1276 psig.