ML20151Z454
| ML20151Z454 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/14/1998 |
| From: | Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20151Z456 | List: |
| References | |
| NUDOCS 9809210344 | |
| Download: ML20151Z454 (13) | |
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PECO NUCLEAR ecco = c -
965 Chesterbrook Boulevard A Unit of PECO Energy wayne. PA 19087-5691
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September 14,1998 1
Docket No. 50-353 i
License No. NPF-85
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U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555
Subject:
Limerick Generating Station, Unit 2 Technical Specifications Change Request No. 98-04-2 i
Reactor Vessel Ma'.erial Surveillance Program l
Dear Sir / Madam:
PECO Energy Company is submittir.? Technical Specifications (TS) Change Request No. 98 2, in accordance with 10CFR50.90, requesting an amendment to the TS (Appendix A) of Facility Operating License No. NPF-85 for Limerick Generating Station (LGS), Unit 2. The proposed changes to the LGS, Unit 2, TS involve revising TS Table 4.4.6.1.3-1. This Table provides the withdrawal schedule for the reactor pressure vessel material surveillance program capsules.
Specifically, this proposed TS change will revise the schedule for withdrawing the first capsule (i.e., Capsule # 131C 7717 G003) from 8 Effective Full Power Years (EFPY) to 15 EFPY, and the second capsule (i.e., Capsule # 131C 7717 G002) from 20 EFPY to 30 EFPY.
l A revision to TS Surveillance Requirement 4.4.6.1.4 is also proposed. This revision will remove the references to flux wire removal and analysis that was originally required following the first cycle of operation. The referenced flux wires were removed following the first cycle of operation, but were misplaced before they were able to be analyzed, as documented in Licensee Event i,
[ f' Report (LER) 2-97-010, dated December 8,1997. This TS Surveillance Requirement will be changed to refer to the flux wires that are located within the surveillance capsules, which will be removed and analyzed in accordance with the surveillance capsule removal schedule located in TS Table 4.4.6.1.3-1.
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k-Q information supporting this TS Change Request is contained in Attachment 1 to this letter, and the proposed TS pages (including marked-up pages) showing the proposed changes to the LGS, Unit 2, TS are contained in Attachment 2. This information is being submitted under affirmation, and the required affidavit is enclosed.
- is a copy of the General Electric (GE) Report No. GE-NE-B1100786-02,
" Surveillance Specimen Program Evaluation for Limerick Generating Station Unit 2," dated June
- 1998, 9809210344 980914 PDR ADOCK 05000353 P
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September 14,1998 :
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We request that, if soproved, the amendment to the LGS, Unit 2. TS be issued by March 31, 1999, and become effective within 30 days of issuance in order to support the upcoming refueling outage (2R05).
If you have any questions, please do not hesitate to contact us.
Very truly yours.
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Garrett D. Edwards.
Director-Licensing Enclosure, Attachments cc:
H. J. Milk.., Jaministrator, Region I, USNRC (w/ enclosure, attachments)
A. L. Burritt, USNRC Senior Resident inspector, LGS (w/ enclosure, attachments)
R. R. Janati, PA Bureau of Radiological Protection (w/ enclosure, attachments) l 1
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1 COMMONWEALTH OF PENNSYLVANIA ss.
COUNTY OF CHESTER J. J. Hagan, being first duly sworn, deposes and says:
l That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the foregoing Application for Amendment to Facility Operating License No. NPF-85 for Limerick Generating Station, Unit 2, concerning Technical Specifications Change Request No. 98-04-2, " Revision to the Surveillance Specimen Removal Schedule," 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.
l x-President Subscribed and swom to before me this /Y day of,
998.
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Notory Public NOTARIAL SEAL QU40L A.WALTON, Notary Pub 8e Csty of Phtledelphia, Phda.Courg h c-ission E4, ires May 28. 200t
ATTACHMENT 1 LIMER!CK GENERATING STATION UNIT 2 DOCKET NO. 50-353 LICENSE NO. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 98-04-2 Supporting information for Change - 9 Pages
" REVISION TO THE SURVEILLANCE SPECIMEN REMOVAL SCHEDULE" h
Dockit No. 50-353 License No. NPF-85 Page 1 of 9 Limerick Generating Station, Unit 2 Technical Specifications Change Request No. 98-04 2 Reactor Vessel Material Surveillance Program Subject PECO Energy Company, licensee under Facility Operating License No. NPF-85 for Limerick Generating Station (LGS), Unit 2, requests that the Technical Specifications (TS) contained in Appendix A to the Operating License be amended as proposed herein, to revise TS Table 4.4.6.1.3-1, " Reactor Vessel Material Surveillance Program - Withdrawal Schedule." This Table provides the schedule for withdrawing the reactor pressure vessel material surveillance program capsules. This proposed TS change involves revising the schedule for withdrawing the first surveillance capsule (i.e., Capsule # 131C 7717 G003) from 8 Effective Full Power Years (EFPY) to 15 EFPY, and the second surveillance capsule (i.e., Capsule #
131C 7717 G002) from 20 EFPY to 30 EFPY.
A revision to TS Surveillance Requirement (SR) 4.4.6.1.4 is also proposed. This revision will remove the references to flux wire removal and analysis that was originally required following the first cycle of operation. The referenced flux wires were removed following the first cycle of operation, but were misplaced before they were able to be analyzed. This condition was reported in Lrxusee Event Report (LER) 2-97-010, dated December 8,1997. TS SR 4.4.6.1.4 will be changed to refer to the flux wires that are located within the surveillance capsules, which will be removed and analyzed in accordance with the surveillance capsulo removal schedule, located in Table 4.4.6.1.3-1.
The proposed changes to the LGS, Unit 2, TS are shown by vertical bars in the margins of the affected TS pages 3/4 4-19 and 3/4 4-21, and are contained in Attachment 2. Marked-up pages indicating the changes are also contained in Attachment 2. of this letter is a copy of General Electric (GE) Report No. GE-NE-B1100786-02,
" Surveillance Specimen Program Evaluation for Limerick Generating Station Unit 2," dated June 1998.
We request that, if approved, the TS changes proposed herein be issued by March 31,1999, and become effective within 30 days of issuance in order to support the upcoming LGS, Unit 2, refueling outage (2R05).
The NRC has previously approved similar TS changes for LGS, Unit 1, by letter dated April 15,1998, for River Bend, Unit 1, by letter dated February 13,1997, and for Grand Gulf Nuclear Station, Unit 1, by letter datec' August 21,1996, based on plant specific circumstances.
This TS Change Request provides a discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, information supporting a finding of No Significant Hazards Consideration, and information supporting an Environmental Assessment.
Discussion and Description of the Proposed Channes The proposed TS Change Request will revise TS Table 4.4.6.1.3-1, " Reactor Vessel Material Surveillance Program - Withdrawal Schedule," for LGS, Unit 2, to change the schedule for withdrawing the reactor pressure vessel material surveillance program capsules. This proposed TS change involves revising the schedule for withdrawing the first surveillance capsule (i.e., Capsule # 131C 7717 G003) from 8 Effective Full Power Years (EFPY) to 15 EFPY, and the second surveillance capsule (i.e., Capsule # 131C 7717 G002) from 20 EFPY to 30 EFPY.
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Dock t No. 50-353 Attachment i License No. NPF-85 Page 2 of 9 A revision to TS Surveillance Requirement (SR) 4.4.6.1.4 is also proposed. This revision will remove the references to flux wire removal and analysis that was originally required following the first cycle of operation. The referenced flux wires were removed following the first cycle of operation, but were misplaced before they were able to be analyzed. This condition was reported in Licensee Event Report (LER) 2-97-010, dated December 8,1997. TS SR 4.4.6.1.4 will be changed to refer to the flux wires that are located within the surveillance capsules, which will be removed and analyzed in accordance with the surveillance capsule removal schedule, located in Table 4.4.6.1.3-1.
Therefore, we propose that LGS, Unit 2, TS Table 4.4.6.1.3-1 and TS SR 4.4.6.1.4 be revised to reflect the proposed changes.
Appendix H of 10 CFR Part 50 requires licensees to withdraw capsules from their reactor vessels periodically according to the capsule withdrawal schedule specified in the American Society for Testing Materials (ASTM) Standard E-185, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." ASTM E-185 provides guidelines for designing a surveillance program, selecting materials, and evaluating test results for light-water cooled nuclear power reactor vessels. Section Ill.B.3 of Appendix H permits alternatives to the recommendations of ASTM E-185, when technically justified and approved by the Nuclear Regulatory Commission (NRC) prior to implementation.
The LG S reactor pressure vessel material surveillance program was designed in accordance with 10CF'.50, Appendix H, and the 1973 edition of ASTM E-185. Appendix H endorses ASTM E-185 and state s that: 'The design of the surveillance program must meet the requirements of the edition of ASTM E-185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased.
Later editions of ASTM E-185 may be used, but including only those editions through 1982.'
LGS is defined as an ASTM E-185-73, Case "A" plant, since the vessel has a predicted shift in the reference nil-ductility temperature (ARTuor) of less than 100'F, and will be exposed to a neutron fluence of less than 5x10" n/cm over the design lifetime of the plant. The current withdrawal schedule specifies the 2
removal of the first and second surveillance capsules at 8 and 20 Effective Full Power Years (EFPY),
respectively. A third capsule is a spare without a specific withdrawal schedule.
If the current schedule for the withdrawal of the first capsule is used, the measured data may not be useful, as the expected shift in RTnor (ARTwor) is small and may be indistinguishable from the data scatter that would typically be experienced from the testing of an unirradiated specimen. The most recently approved ASTM E-185 guidance (ASTM E-185-82) regarding first capsule withdrawal states: "The first capsule is scheduled for withdrawal early in the vessellife to verify the initialpredictions of the surveillance materialresponse to the actual radiation environment. It is removed when the predicted shift exceeds the expected scatter by sufficient margin to be measurable." Since the LGS, Unit 2, vessel material expected shift is low, the first surveluance capsule testing should be deferred to when the majority of the shift in the vessel RTuor has been achieved and is expected to be measurable. Removal and testing at the revised withdrawal schedule will permit the collection of more credible data for fracture toughness predictions.
Therefore, the surveillance program's first capsule withdrawal schedule should be extended from 8 EFPY to 15 EFPY. Also, in order to collect useful and meaningful data from the second capsule, its withdrawal schedule should be extended from 20 EFPY to 30 EFPY. This will permit adequate additional exposure following the removal and analysis of the first capsule, as well as permitting a withdrawal schadule that is consistent with the withdrawal schedule as approved for Limerick Generating Station Unit 1. In addition to withdrawing the specimen when the predicted shift exceeds the expected scatter by sufficient margin to be measurable, the revised schedule for capsule withdrawal continues to meet the :ntent of ASTM E185-82, as the first capsule will be removed with the fluence less than 5x10* n/cm and t.ie value of ARTwor less 2
than 50'F, and the second capsule will be removed before the accumulated neutron fluence of the capsule corresponds to the approximate end of life (EOL) fluence at the reactor inner wall location.
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l Dock:t No. 50-353 Attachment i License No. NPF-85 Page 3 of 9 Safety Assessment The original surveillance capsule withdrawal schedule was developed in accordance with the intent of 10CFR50, Appendix H. This schedule did not account for the following LGS specific conditions:
Good plate and weld chemistry (Iow copper content from 0.01% - 0.15%);
Low RPV 1/4T 32 EFPY beltline fluence ( 5x10 n/cm )
2 Resulting low predicted shift in the reference nil-ductility temperature, RTuor (<60*F at 32 EFPY).
. contains a technical report prepared by GE (i.e., GE Report No. GE-NE-B1100786-02,
" Surveillance Specimen Program Evaluation for Limerick Generating Station Unit 2"). This report provides the details associated with the rpecific conditions that exist at LGS, and provides the basis for revising the withdrawal schedule for the first surveillance capsule to be removed at 15 EFPY and the second surveillance capsule to be removed at 30 EFPY. The withdrawal schedule for the third capsule (spare) remains unchanged.
The first surveillance capsule testing schedule should be developed to measure a significant portion of the fracture toughness change, as measured by ARTuor. Since the limiting plate material for the LGS, Unit 2, reactor pressure vessel has a low expected ARTwor of 48 F (at 1/4T) over the life of the plant, the recommended schedu!e should be designed to measure a majority of ARTuor of the plate material. Given the low expected shift, a criteria of 75% of the expected shift in RTwo7was selected to determine the revised schedule. For LGS, Unit 2,75% of the expected shift is 0.75 (48) = 36*F. Using a criteria of 75%
of the expected shift (36*F), the capsub will experience this shift for the plate material at approximately 29 EFPY. The removal of the first capsulc
- conservatively set at 15 EFPY to provide consistency with LGS, Unit 1. In support of this schedule, at 15 EFPY, an expected 26 F shift will have occurred in the surveillance capsule limiting material, which will provide sufficient data to determine the required vessel material properties.
As specified in Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, Regulatory Position 2, " Surveillance Data Available,"the collection of two (2) or more sats of credible surveillance data is necessary to empirically calculate the adjusted reference temperature (ART). Each surveillance capsule constitutes only one (1) set of credible surveillance data. This calculated ART can be used to revise the Pressure-Temperature (P-T) curves (i.e., TS Figure 3.4.6.1-1). Without two (2) or more sets of credible data, the ART must be calculated and the P-T curves revised, based upon the calculational methodologies delineated in the Regulatory Guide 1.99, Revision 2. Regulatory Position 1,
" Surveillance Data Not Available." These methodologies use plant specific chemistry and fluence values to determine a calculated shift in RTwor. A " margin" term is then added, to obtain conservative, upper-bound values of adjusted reference temperature.
The existing LGS, Unit 2, P-T curves are based upon the Regulatory Position 1 methodology, and are currently valid up to 10 EFPY. With first capsule removal at either 8 o_r 15 EFPY, the existing P-T curves will require a revision, prior to reaching 10 EFPY, based upon the calculational methodologies as contained in the Regulatory Guide 1.99, Revision 2, Regulatory Position 1,
- Surveillance Data Not Available." Therefore, the TS revision to the first capsule withdrawal schedule results in no impact to the calculational methodologies that will be used for the P-T curve revision that will be necessary to extend me cuNes beyond 10 EFPY.
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1 Dock:t No. 50-353 License No. NPF-85 Page 4 of 9 As specified in the report contained in Attachment 3, justification for the revision of the first capsule withdrawal schedule from 8 EFPY to 15 EFPY is based on the following:
1.
Based on ART calculations performed in accordance with Regulatory Guide 1.99, Revision 2, the shift (ARTer + margin) for the LGS, Unit 2, surveillance plate is calculated to be 68*F at 32 EFPY. If the first capsule is removed at 8 EFPY, the actual shift (predicted to be 19 F) may not be large enough to t.e differentiated from the data scatter, since the predicted fluence on the capsule at 8 EFPY (3.9x10" 2
n/cm ) is low, and the chemistry of the LGS, Unit 2, reactor pressure vessel plate material is good (0.11%-0.15% copper). Thus, the data obtained may not be useful for predicting the material behavior, as it may not be distinguishable from the unirradiated data. Therefois, removal of the i
capsule at 15 EFPY satisfies the intent of ASTM E-185, in that the predicted shift is expected to exceed the expected scatter by a sufficient margin to be detectable.
2.
Based on a review of predicted RTm1 shifts and measured RTuor shifts from other Boiling Water Reactor (BWR) surveillance capsules, the predicted shifts bound the measured results. Figure 2-1 of j is a plot of actual shift measurements versus predicted shifts (calculated per Regulatory Guide 1.99, Revision 2) for base material. This figure shows that the predicted shift plus margin conservatively bounds the actual shifts measured from BWR surveillance specimen data. The same plot for weld material (Figure 2-2) again shows the predicted shift plus margin term bounds the measured shift.
3.
The LGS fluence used for shift predictions in accordance with Regulatory Guide 1.99, Revision 2 is based upon a conservative calculation, and is expected to bound the actual fluence.
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The fluence data as determined from the surveillance capsule flux wires at 15 EFPY will provide an accurate indication of neutron fluence. In accordance with Regulatory Guide 1.99, Revision 2, Regulatory Position 1 methodology, data from these flux wires will permit an adjustment of TS Figure 3.4.6.1-1 in accordance with surveillance requirement 4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50, Appendix H, and ASTM E-185.
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The Supplemental Surveillance Program (SSP) is a BWR Owner's Group (BWROG) program, designed to increase the amount of surveillance data in a systematic manner. As part of this program, the BWROG prepared supplemental capsules which were installed in Cooper and Oyster Creek. The j
SSP specimens will provide early test data for vessel plate, which is similar to the LGS, Unit 2, surveillance plate. The SSP will supplement the LGS, Unit 2, surveillance program by providing timely detection of anomalous RTwor shifts, should any occur.
The combination of the low expected shift for the reactor pressure vessel plate material, SSP data on similar material, and the inherent margin in the P-T curve calculations will result in a credible set of surveillance data, while ensuring the continued safe operation of LGS, Unit 2.
Also, as specified in the Attachment 3, it is appropriate to extend the withdrawal of the LGS, Unit 2, second capsule. The current schedule specifies withdrawal of the second capsule at 20 EFPY. Based upon the information provided in Attachment 3 supporting withdrawal of the first capsule at 15 EFPY, there I
will be an insignificant shift in material properties at 20 EFPY, after only an additional exposure of 5 EFPY, It is appropriate to extend this schedule to 30 EFPY which meets the intent of ASTM E185-82, such that the withdrawal of the second capsule occurs before the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the reactor inner wall location, and provides consistency with the Limerick, Unit 1, withdrawal schedule.
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Docket No. 50-353 License No. NPF-85 Page 5 of 9 A revision to TS SR 4.4.6.1.4 is also proposed. This proposed revision changes the reference to the flux wires to be analyzed from "at the first refueling outage" to the flux wires located within the surveillance capsules which will be analyzed in accordance with TS Table 4.4.6.1.3-1. This revision is necessary to support the update of the surveillance requirements to reflect the elimination of the use of first cycle flux data, since this data was not collected, and the addition of the use of flux data from the flux wires located within the surveillance capsules. LGS LER 2-97-010 provides additional information concerning the failure to perform a first cycle analysis of a reactor flux wire specimen.
Information Supportina a Findina of No Sianificant Hazards Consideration We have concluded that the proposed changes to the Limerick Generating Station (LGS), Unit 2, Technical Specifications (TS) which will revise TS Table 4.4.6.1.3-1 to change the schedule for withdrawing the first and second surveillance capsules from 8 Effective Full Power Years (EFPY) to 15 EFPY and from 20 EFPY to 30 EFPY, respectively; and TS Surveillance Requirement (SR) 4.4.6.1.4 to remove the reference to the flux wire removal and analysis that was originally required, do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10CFR50.92 is provided below:
1.
The proposed Technical Specifications (TS) chanaes do not involve a sianificant increase in the probability or consecuences of an accident previously evaluated.
The proposed changes do not increase the probability of occurrence of an accident previously evaluated in the safety analysis report and do not affect any accident initiators as described in the Safety Analysis Report (SAR). The change revises the withdrawal schedule for the reactor vessel material surveillance capsules. The capsules are not an initiator of any previously analyzed accident nor does the withdrawal schedule of the surveillance capsules affect the probability or consequences of any previously analyzed accident.
The proposed changes will not affect the Pressure-Temperature (P-T) limits as specified in LGS TS Figure 3.4.6.1-1 and Updated Final Safety Analysis Report (UFSAR) Figure 5.3-4. P-T limits are imposed on the reactor coolant system to ensure that adequate safety margins exist during normal operation, anticipated operational occurrences, and system hydrostatic tests. The P-T limits are related to the RTuo7, as described in ASME Section 111, Appendix G. Changes in the fracture toughness properties of RPV beltline materials, resulting from neutron irradiation and the thermal environment, are monitored by a surveillance program in compliance with the requirements of 10CFR50 Appendix H. The effect of neutron fluence on the shift in the RTuor is predicted by methods given in Regulatory Guide 1.99, Rev.2.
As detailed in Attachment 3, for LGS, Unit 2, the combination of low expected RTuor shift for the plate material due to low predicted fluence and excellent material chemistry; Supplemental Surveillance Program (SSP) data on similar material; and the inherent margin in the P-T curve calculations, with the withdrawal schedule of the first surveillance capsule modified from 8 EFPY to 15 EFPY and the second surveillance capsule modified from 20 EFPY to 30 EFPY, will result in more credible sets of surveillance data, while ensuring the continued safe operation of LGS, Unit 2.
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Dock:t No. 50-353 Attachment i License No. NPF-85 Page 6 of 9 The current LGS P-T limits were established based on adjusted reference temperatures developed in accordance with the procedures prescribed in Regulatory Guide 1.99, Revisien 2 Regulatory Position 1,
- Surveillance Data Not Available." Calculation of adjusted reference temperature by these procedures includes a conservative base fluence estimate; power rerate adjustment of a 110% fluence multiplier from startup, instead of a 105% fluence multiplier since 2R03; and a margin term to ensure conservative, upper-bound values are used for the calculation of the P-T limits. Revision of the first capsule withdrawal schedule will not affect the P-T limits because they will continue to be established in accordance with Regulatory Position 1 guidance.
1 Also, as indicated in Attachment 3, it is also appropriate to extend the withdrawal of the LGS, Unit 2, second capsule. The current schedule specifies withdrawal of the second capsule at 20 EFPY.
Based upon the information provided in Attachment 3 supporting withdrawal of the first capsule at 15 EFPY, there will be an insignificant shift in material properties at 20 EFPY, after only an additional exposure of 5 EFPY. It is appropriate to extend this schedule to 30 EFPY which meets the intent of ASTM E185-82, such that the withdrawal of the second capsule occurs before the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the reactor pressure vessel inner wall location, and provides consistency with the LGS, Unit 1, withdrawal schedule.
in accordance with the guidancs stipulated in Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, Regulatory Position 2, " Surveillance Data Available,"the i
collection of two (2) or more sets of credible surveillance data is necessary to empirically calculate the adjusted reference temperature (ART). Each surveillance capsule constitutes one set of credible surveillance data. This calculated ART can be used to revise the P-T curves (TS Figure 3.4.6.1-1). Without two (2) or more sets of credible data, the ART must be calculated and the P-T curves revised, based upon the calculational methodologies as provided in the Regulatory Guide 1.99, Revision 2, Regulatory Position 1, " Surveillance Data Not Available." These methodologies use plant specific chemistry and fluence values to determine a calculated shift in RTuo7 A
" margin" term is then added, to obtain conservative, upper-bound values of adjusted reference temperature.
The existing LGS, Unit 2, P-T curves are based upon the Regulatory Position 1 methodology, and are currently valid up to 10 EFPY. With first capsule removal at either 8 or 15 EFPY, the existing P-T curves will require a revision, prior to reaching 10 EFPY, based upon the calculational methodologies as contained in the Regulatory Guide 1.99, Revision 2, Regulatory Position 1, "Suiveillance Data Not Available." Therefore, the Technical Specification revision to the first i
capsule withdrawal schedule, as supported by this Safety Evaluation, results in no impact to the calculational methodologies that will be used for the P-T curve revision that will be necessary to extend the curves beyond 10 EFPY.
The fluence data as determined from the surveillance capsule flux wires at 15 EFPY will provide i
an accurate indication of neutron fluence. In accordance with Regulatory Guide 1.99, Revision 2, Regulatory Position 1 methodology, data from these flux wires will permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR 4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50, Appendix H, and ASTM E-185.
The proposed changes will not affect any plant safety limits or limiting conditions of operation.
The proposed changes will not affect reactor pressure vessel performance as it involves no physical changes and LGS P-T limits will remain conservative in accordance with Regulatory Guide 1.99, Revision 2, guidance. The proposed changes will not cause the reactor pressure
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vessel or interfacing systems to be operated outside of their design or testing limits.
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License No. NPF-85 Page 7 of 9 i
1 The proposed changes do not increase the probability of the occurrence of a malfunction, or consequences of a malfunction, of equipment important to safety previously evaluated in the SAR.
The proposed changes do not involve any physical changes to equipment important to safety.
The potential for reactor vessel failure will be adequately assessed by the proposed withdrawal schedule. In addition, the results from the Supplemental Surveillance Program (SSP) will provide industry data that bounds the materials used in the LGS vessel until the data from the first LGS capsule is available. The proposed change provides the same level of confidence in the integrity of the vessel. The P-T curves are currently controlled by the TS and are determined using the conservative methodology delineated in Regulatory Guide 1.99. Therefore, the possibility of failure of the reactor vessel is not increased. The current P-T limit curves are inherently conservative and will continue to be adhered to.
Therefore, the proposed TS changes do not involve an increase in the probability or consequences of an accident previously evaluated.
2.
The proposed TS chanaes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a different type of accident than any previously evaluated in the SAR. The proposed changes are a revision of the withdrawal schedule for the first reactor pressure vessel material surveillance capsule from 8 EFPY to 15 EFPY, and for the second capsule from 20 EFPY to 30 EFPY. The proposed changes do not involve a physical modification of the design of plant structures, systems, or components. The proposed changes will not impact the manner in which the plant is operated as plant operating and testing procedures will not be affected by the change. No new accident types or failure modes will be introduced as a result of the proposed change.
LGS's current P-T limits were established based on adjusted reference temperatures developed in l
accordance with the procedures prescribed in Regulatory Guide 1.99, Revision 2, Regulatory Position 1. " Surveillance Data Not Available." Calculation of adjusted reference temperature by these procedures includes a conservative base fluence estimate; power rerate adjustment of a 110% fluence multiplier from startup, instead of a 105% fluence multiplier since 2R03; and a l
margin term to ensure conservative, upper-bound values are used for the calculation of the P-T limits. Revision of the first capsule withdrawal schedule will not affect the P-T limits because they will continue to be established in accordance with the guidance of Regulatory Position 1 of Regulatory Guide 1.99. Also, as specified in Attachment 3, it is appropriate to extend the withdrawal of the LGS, Unit 2, second capsule. The current schedule specifies withdrawal of the second capsule at 20 EFPY. Based upon the information provided in Attachment 3 supporting withdrawal of the first capsule at 15 EFPY, there will be an insignificant shift in material properties at 20 EFPY, after only an additional exposure of 5 EFPY. It is appropriate to extend this schedule to 30 EFPY which meets the intent of ASTM E185-82, such that the withdrawal of the second capsule occurs before the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the reactor inner wall location, and provides consistency with the LGS, Unit 1, withdrawal schedule.
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Dock:t No. 50-353 Attachment i License No. NPF-85 Page 8 of 9 1
The existing LGS, Unit 2, P-T curves are based upon the Regulatory Position 1 methodology, and are currently valid up to 10 EFPY. With first capsule removal at either 8 or 15 EFPY, the existing P-T curves will require a revision, prior to reaching 10 EFPY, based upon the calculational methodologies as contained in the Regulatory Guide 1.99, Revision 2, Regulatory Position 1,
" Surveillance Data Not Available." Therefore, the proposed TS revision to the first capsule withdrawal schedule results in no impact to the calculational methodologies that will be used for the P-T curve revision that will be necessary to extend the curves beyond 10 EFPY.
The fluence data as determined from the surveillance capsule flux wires at 15 EFPY will provide an accurate indication of neutron fluence. In accordance with Regulatory Guide 1.99, Revision 2, Regulatory Position 1 methodology, data from these flux wires will permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR 4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50, Appendix H, and ASTM E-185.
l The potential for reactor vessel failure will be adequately assessed by the proposed withdrawal schedule. In addition, the results from the SSP will provide industry data that bounds the materials used in the LGS vessel, until the data from the first LGS capsule is availablo. The proposed changes provide the sarre level of confidence in the integrity of the vessel. The P-T curves are currently controlled by the TS and are determined using the conservative methodology in Regulatory Guide 1.99. Therefore, the possibility of failure of the reactor vessel is not increased. The current P-T limit curves are inherently conservative and will continue to be adhered to.
Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
The proposed TS chances do not involve a sianificant reduction in a marain of safety.
The proposed changes to the TS do not reduce the margin of safety as defined in the Bases for any TS. The proposed changes will not affect any safety limits, limiting safety system settings, or limiting conditions of operation. The proposed changes do not represent a change in initial conditions, system response time, or in any other parameter affecting the course of an accident analysis supporting the Bases of any TS. The proposed changes do not involve revision of the P-T limits, but rather a revision of the withdrawal schedule for the surveillance capsules. The current P-T limits were established based on the adjusted reference temperatures for reactor pressure vessel beltline materials calculated in accordance with the guidance stipulated in Regulatory Position 1 of Regulatoq Guide 1.99, Revision 2. P-T limits will continue to be revised as necessary for changes in adjusted reference temperature due to changes in fluence according to Regulatory Position 1 until two (2) or more credible surveillance data sets becomes available.
When two (2) or more credible surveillance data sets become available, P-T limits will be revised as prescribed by Regulatory Position 2 of Regulatory Guice 1.99, Revision 2, or other NRC approved guidance.
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l Docket Ns. 50-353 l
License No. NPF-85 Page 9 of 9 The current P-T limit curves are inherently conservative and provide sufficient margin to ensure the integrity of the reactor vessel.. The changes do not adversely affect these curves. The fluence data as determined from the surveillance capsule flux wires at 15 EFPY will provide an accurate indication of neutron fluence. In accordance with Regulatory Guide 1.99, Revision 2, Regulatory Position 1 methodology, data from these flux wires will permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR 4.4.6.1.3, if required, and will meet the requirements of -
10 CFR 50, Appendix H, and ASTM E-185.
Therefore, the proposed TS changes do not involve a reduction in a margin of safety, information Supportina an Environmental.A-rament An environmental assessment is not required for the changes proposed by this TS Change Request because the requested changes to the Limerick Generating Station (LGS), Unit 2, Technical Specifications (TS) conform to the criteria for " actions eligible for categorical exclusion" as specified in 10CFR51.22(c)(9). The requested changes will have no impact on the environment. The proposed changes do not involve a significant hazards consideration as discussed in the preceding section. The proposed changes do not involve a significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Conclusion The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the Limerick Generating Station (LGS), Unit 2, Technical Specifications (TS) and have concluded that they do not involve an unreviewed safety question, and will not endanger the health and
- safety of the public.
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