ML20058C564
| ML20058C564 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/17/1993 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19312B493 | List: |
| References | |
| NUDOCS 9312020490 | |
| Download: ML20058C564 (7) | |
Text
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. jl PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBROOK BLVD.
, j WAYNE, PA 19087-5 131 l
(215) 640-6000
. k November 17, 1993
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h STATION SUPPORT oEPAnTMENT
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Docket Nos. 50-352 50-353
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License Nos. NPF-39 i
NPF-85 g
U. S. Nuclear Regulatory Cormission
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A1.tn: Document Control Desk 4
Washington, DC 20555
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SUBJECT:
Limerick Generating Station, Units 1 and 2
- E Technical Specifications Change Request 92-08-0 q
Additional Information j
l Gentlemen:
On August 27, 1993, Philadelphia Electric Company (PECo) submitted 3
Technical Specifications (TS) Change Request No. 92-08-0 to the NRC, in j
accordance with 10 CFR 50.90, requssting an amendment to the TS (i.e., Appendix i
A) of Operating License Nos. NFF-39 and NPF-SS for Limerick Generating Station 2
(LGS), Units 1 and 2, respectively. The proposed TS changes reflect an expanded l3 operating domain for LGS, Units 1 and 2,
resulting from the proposed
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implementation of the Average Power Range Monitor - Rod Block Monitor Technical Specifications /Maxicium Extended Load Line Limit Analysis (ARTS /MEl.LLA).
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- q As a result of additional reviews of the proposed TS changes within PEco, d
we have identified minor technical and typographical corrections, described ll belcw, to the changes proposed in our August 27, 1993 letter.
This letter i i provides the proposed TS change pages incorporating those corrections.
As a result of the technical corrections, General Electric (GE) has revised
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the safety analysis report, " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units I and 2,"
a NEDC-32193P, that supported TS Change Request 92-08-0, and was submitted to the i j NRC by our letter dated August 27, 1993. Specifically, NEDC-32193P, Revision 2, 4 g dated October 1993, incorporates the technical corrections to Revision 1 of NEDC-a 32193P with respect to the Average Power Range Monitor (APRM) flow-biased t introl 3!
rod block trip setpoints and allowable values.
Accordingly, the associated corrections to the following proposed TS pages included in our August 27, 1993 letter are described blow. These corrections apply to both LGS Unit I and Unit l
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t U. S. Nuclear R:gulatory Commissicn November 17, 1993 Document Centr 01 Desk Page 2 k
1)
TS Table 2.2.1-1, " Reactor Protection System Instrumentation Setpoints,"
TS page 2-4.
New nominal trip setpoints and allowable values were established for the APRM flow biased high flux SCRAM by using a more restrictive calibration tolerance than that used in the original calculations. This adjustment was done in accordance with a request from LGS personnel. The analytical limit remains at 121%.
2)
TS Table 3.3.6-2, " Control Rod Block Instrumentation Setpoints," TS page 3/4 3-60.
The previout values were calculated by GE assuming a 108%
analytical limit. This resulted in overly conservatee allowable values that would have restricted plant operation.
The APRM control rod block analytical limit and allowable value were recalculated by GE based on a nominal trip setpoint of 108%, which is standard industry practice. The methodology applied has not changed; however, due to the change in input the results are different and provide more flexibility for plant operation.
I We have reviewed the above described technical corrections incorporated into the safety analysis report " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2,"
NEDC-32193P, Revision 2, dated October,1993, and the associated changes to the proposed 1S pages, against the information provided in our letter to the NRC, dated August 27, 1993 (i.e., the Safety Assessment, the information supporting a finding of No Significant Hazards Consideration, and the information supporting an Environmental Assessment), and L have determined that these changes do not alter the statements and conclusions documented in that letter.
The TS pages that contained typographical errors have been corrected and are identified by an asterisk on the Attachment I cover sheet.
These typographit.al changes do not alter any of the statements provided in our letter to the NRC, dated August 27, 1993.
Attachment I contains a complete set of corrected proposed TS pages which supersede the TS pages provided in our lettar to the NRC, dated August 27, 1993.,
" Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Gcaerating Station Units 1 and 2,"
NEDC-32193P, Revision 2, dated October,1993, contains the corrected safety analysis prepared by GE to support the implementation of ARTS /MELLLA at LGS, Units 1 and 2, and reflects the changes to the APRM flow-biased control rod block trip setpoints and allowable values.
Pages 2-4, 3-2, and 11-1 of the attached NEDC-32193P incorporate the corrections discussed above and the specific changes are indicated by change brackets in the page margin. contains information proprietary to GE. GE requests that the document be withheld from public disclosure in accordance with 10 CFR 2.790(6)(4).
The affidavit supporting this request in accordance with 10 CFR 2.790(b)(1), is provided with Attachment 2.
s U. S. Nuclear R:gulatory Commission November 17,-1993 s
Document Control Desk Page 3 This letter also provides the following responses to requests for additional information that were discussed between Mr. Jonathan Winter of the NRC-l and R. Krich of PEco on October 19, 1993.
Each question has been restated followed by our response.
Question No. 1:
Does the 42% recirculation pump runback intermediate speed setpoint for the recirculation pump speed, keep plant i
operation outside the instability region?
l I
Response
For operation along the Maximum Extended Load Line Limit (MELLL) rod-l line at LGS, GE has recommended that the recirculation pump rurback-intermediate speed setpoint be reset from the current 47% value to the new 42% value. This lower pump speed setpoint would correspond to about 81% core power / 54% core flow operation along the MELLL rod line, and 69% core power /56% core flow along the rated rod line.
There is adequate margin between the new minimum core flow condition and the 45% core flow boundary of the stability exclusion region ~.
l Ouestion No. 2t What is the sensitivity of the suppression pool temperature to boron (B)-10 enrichment?
Response
While no specific sensitivity calculations have been performed for LGS, a sensitivity study for the Boiling Water Reactor (BWR) type 4 product line was performed as part of the GE report, " Assessment of BWR Mitigation of ATWS," NEDE-24222, dated February 1981.
This study showed the sensitivity of the suppression pool temperature to boron injection rate. For a 50% reduction in injection flow rate, the suppression pool temperature increased by approximate 1y'48' F.
This reduction in flow rate results in approximately the same reduction in injection rate of B-10 enriched boron, as reducing the B-10 enrichment from 29% to natural (i.e.,19.9%) enrichment at a constant flow rate.
This would conservatively ' indicate less than 5*F change in suppression pool temperature for each 1% change in B-10 enrichment.
Question No. 3:
What is the value of 6 PCT, used in the SAFER /GESTR analyses as described on page 3-2 of the " Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32170P, Revision 1, dated June, 19937 Resoonse:
The value of 6 PCT, used in the LGS SAFER /GESTR analyses, is 46.7*F.
4 U. S. Nuclear R:gulatory Commission Nov:mber 17, 1993 Document Control Desk Page 4 Ouestion No. 4:
Why there is a dip on Figure 5-1 of
" Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32170P, Revision 1, dated June,1993?
Response
The minimum Peak Cladding Temperature (PCT) point in the dip of Figure 5-1 corresponds to a break size of 0.5 ft.
This break size is not affected as strongly as smaller breaks to the relaxation in the Emergency Core Cooling Systems (ECCS) performance parameters.
The duration of core uncovery is relatively short in this case, resulting in a small rise in PCT, which is quickl{ terminated by the ECCS injection flow.
In contrast, the 0.07 ft break (i.e., the limiting small break) is very sensitive to the ECCS parameter relaxation.
The ECCS injection becomes significantly delayed, allowing a longer core uncovery period to occur. Accordingly, the resulting rise in PCT is larger than for the 0.5 ft" break for this reason.
Question No. 5 Explain the values of the Minimum Critical Power Ratio (MCPR) for the Feedwater Controller failure (FWCF) transient at 100%
power and 105% core flow as shown in Table 5-3 of " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," NEDC-32193P, Revision 1, dated July, 1993.
Response
The MCPR limit chosen for the verification of the generic portion of the MCPR(P) curve was 1.34, based on the Turbine Trip with No Bypass (TTNBP) event at rated conditions.
This limit is reasonable for confirming the MCPR(P) curve, provided that the Recirculation Pump Trip (RPT) and turbine bypass valves are operable.
Under the condition cf RPT and turbine bypass valves inoperable, the MCPR limit can be considerably higher, in the range of 1.43.
The analyses presented in Table 5-6 of NEDC-32193P are based on RPT and turbine bypass valves operable only. A separate set of analyses has been performed by GE for the RPT and/or turbine bypass valves inoperable option.
These analyses were not included in the document, due to schedule constraints. The plant specific portion of the MCPR(P) curve is also affected by inoperable RPT and/or turbine bypass valves. The Core Operating Limits Report (COLR) will include MCPR(P) curves for both the conditions of operable and inoperable RPT and/or turbine bypass valves.
Question No. 6 Rephrase statement on TS Bases page B 3/4 2-4, related to the power dependent MCPR limits, to better explain the current wording " bypassed power level."
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U. S. Nuclear R:gulatory Commission Novnber 17, 1993 Document Control Desk Page 5
Response
The statement on TS Bases page B 3/4 2-4 has been revised to read
... operating limits are provided for operating between 25% RTP and 30% RTP."
Ouestion No. 7 How does the power dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) multiplier, (i.e., MAPFAC), get into the process computer?
Response
The process computer code is ARTS /MELLLA compatible, therefore GE will input MAPFAC into the process computer data base, as part of the ARTS /MELLLA implementation activities.
Question No. 8 The ARTS /MELLLA TS changes for TS Table 3.3.6-2, " Control Rod Block Instrumentation _ Setpoints," TS page 3/4 3-60, do not include clamped trip setpoints for the APRM flow biased high flux scram. Why not?
Resoonse The clamped trip setpoints are not part of the current TS for the APRM flow biased high flux scram at LGS, therefore they are not provided for the proposed ARIS/MELLLA TS changes.
Question No. 9 TS Table 3.3.6-2,
" Control Rod Block Instrumentation Setpoints, " TS page 3/4 3-60, shows different values for the Rod Block Monitor - power range setpoint (i.e., for the Low Power Setpoint, the range is between 23% and 26% rated thermal power), than the. range specified in the TS Section 1.0, " DEFINITIONS," TS page 1-4 (i.e., for Low Power Trip Setpoint, the range is between 30% and 65% reactor thermal power).
Response
The information on both TS pages is correct. The range specified on TS page 1-4 refers to analytical limits, while the range specified on TS page 3/4 3-60 refers to actual nominal-trip setpoints and allowable values.
We request that, if approved, the amendments be issued on January 19, 1994, for LGS Unit 1, and February 18, 1995, for LGS Unit 2 and be made effective within 30 days after the issue date. This is necessary in order to allow testing of the Standby Liquid Control System (SLCS) operation with enriched boron. We will notify you of the actual dates when the amendments are put into effect.
These dates support the planned modification schedules.
Specifically, the modifications associated with implementation of the ARTS /MELLLA changes are planned to be made during the fifth refueling outage for Unit 1, currently scheduled to begin on January 29, 1994, and the third refueling outage for Unit 2, currently scheduled to begin on March 4, 1995.
U. S. Nuclear Regulatory Commission t
Document Control Desk November 17, 1993 Page 6 If you have any questions, please do not hesitate to contact us.
Very truly yours,
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G.
. Hunger, Jr.
Director Licensing Section Attachments T. T. Martin, Administrator, Region I, USNRC w/ attachments cc:
N. S. Perry, USNRC Senior Resident Inspector, LG.i W. P. Dornsife, Director, PA Bureau of Radiological Protection w/ attachments i
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i COMMONWEALTH OF PENNSYLVANIA:
ss.
COUNTY OF CHESTER G. R. Rainey, being first duly sworn, deposes and says:
That he is Vice President of Philadelphia Electric Company; the Applicant herein; that he has read the corrected information for Technical Specifications Change Request No. 92-08-0 to reflect the -implementation of the Average Power Range Monitor - Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA) at Limerick Generating Station Units 1 and 2, 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.
s dat^r M g si Vice President Subscribed and sworn to before me this day
-ffcI}h/JAO>[4 1993.
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s Notary Public Notat! Seal Tw--J%
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