ML20198M786
| ML20198M786 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 01/12/1998 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20198M758 | List: |
| References | |
| NUDOCS 9801200120 | |
| Download: ML20198M786 (14) | |
Text
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ATTACHMENT 1 LIMERICK GENERATING STATION UNIT 1 4
DOCKET NO. 50 352 LICENSE NO. NPF-39 i--
TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 97-02-1 l
l l-Supporting Information for Change - 9 Pages
" REVISION TO THE SURVEILLANCE SPECIMEN -
REMOVAL SCHEDULE" r
9 9901200120 990112 PDR ADOCK 05000352 V
P PDR
a Dock 3t No. 50-352 License N3. NPF-39 PECO Energy Company, under Facility Operating License No. NPF;39 for Limerick Generating Station (LGS), Unit I 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as proposed herein, to revise TS Table 4.4.6.1.3-1 to change the withdrawal-schedule for the first capsule to be withdrawn (Capsule _# 117C 4944 G004) from 10 Effective full Power Years (EFPY) to 15 EFPY.
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 never located following the first cycle of operation (LER 87-032-00). This 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.
The proposed changes to the LGS, Unit 1 TS are indicated by markups on TS pages 3/4 4-19 and 3/4 4-21.
The TS pages showing the proposed changes are contained in Attachment 2. is-a copy of the GE Report No. GE-NE-B1100786-01R1, " Surveillance Specimen Program Evaluation for Limerick Generating Station, Unit 1," dated December 1997.
We request that, if approved, the TS changes proposed herein be issued by March 31, 1998, and become effective within 30 days of issuance in order to support the upcoming refueling outage (IR07).
The NRC has previously approved similar TS changes at River Bend Station, Unit 1 by letter dated February 13, 1997, Lnd at Grand Gulf Nuclear Station, Unit I by letter dated 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 change::, information supporting a finding of No Significant Hazards Consideration, and information supporting an Environmental Assessment.
Discussion and Description of the Prqoosed Chad 9M The proposed Technical Specifications (TS) Change Request will revise Limerick Generating Station (LGS), Unit 1 TS Table 4.4.6.1.3-1 to change the withdrawal schedule for the first capsule to be withdrawn (Capsule # 117C 4944 G004) from 10 Effective Full Power Years (EFPY) to 15 EFPY.
A revision to TS Surveillance Requirement 4.4.6.1.4 is also proposed.
This revision will remove the references to flu:: wire removal and analysis that was originally required following the first cycle of operation. The referenced flux wires were never located following the fIrst cycle of operation (LER 87-
'032-00). This Surveillance Requirement will be changed to refer to the flux j
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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Docket-No.L50-352-T, License ' Na. : NPF-39_.
s L Therefore', we propose ~ that'. LGS Unit 1, TS table.4.4.6.1.3-11 andiTS '
L urveillance Requirement:4.4.6.1.4 be revised to reflect these changes.
S L Appendix / fo^f:10 CFR:Part 50 requires. licensees to_ withdraw; capsules: from?
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their reactor vessels periodically according to theAapsule withdrawal
. schedule specified:in the American Society for Testing Materials (ASTM):
1
- 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 III.B.3 of Appendix H permits alternatives to the-recommendationsLof 2 e
. ASTM E-185,: when technically justified and approved by the Nuc16ar Regulatory t
Commission (NRC) prior'to implementation.
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2 The.-LGS: reactor pressure vessel material: surveillance program was designed in -
.accordance with 10CFR50, Appendix H, and-the 1973-edition of' ASTM E-185.
- Appendix.H endorses E-185 and states that "The design of the-surveillance-j program must meet the requirements of the edition of ASTM E-1851that--is -
[
l 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
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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 (ART,.,) of less -
than 100*F and will be exposed to -a neutron fluence of less than 5x10 n/ca' over the design lifetime of.the plant. The current withdrawal schedule
. specifies the removal of the first and second surveillance capsules at 10 and withdrawal schedule. ~ A third capsule is a spare without a specific 30 EFPY, respectively.
q If the current schedule for the withdrawal of the first capsule is used, the L
. measured data 'may not be 'useful, as the expected shift in RT,., (ART,.,) is-
, small and may be-indistinguishable from the data scatter that would typically p
be experienced from the testing of an um eradiated specimen. The most E
_ recently approved ASTM E-185 guidance (A TM E-185-82) regarding first capsule withdrawal states:
"The first capsule is :cheduled for withdrawal early in
. the vessel life to verify the initial pred'.ctions of the surveillance material response 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 I vessel material expected shift -is low, the N
first surveillance capsule-testing should be deferred to when the majority of
- the shift in the vessel RT,,, has been achieved and is expected to be r
' measurable. Removal and testing at the revised withdrawal schedule will opermitithe collection of more credible data for fracture toughness i
predictions. Therefore, the surveillance program's first capsule withdrawal
-schedule should be extended from 10 EFPY to 15 EFPY.- In-addition to (withdrawing the specimen when=the predicted shift-exceeds the expected scatter b'
. by sufficient margin to be measurable, the.. revised schedule for-the withdrawal of(the first capsule continues to meet the intent of ASTM E-185-82, as the first capsule will be removed.with the fluence less than 5 x'10 n/cm' and the 1 value - of ' ART,,,' less than 50"F.
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Dock;t No. 50-352 4
s License No, NPF-39
' 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 Limerick Generating Station (LGS), Unit I specific conditions:
Excellent plate and weld chemistry (low copper of 0.01-0.12%);
Low RPV 1/4T 32 EFPY beltline fluence ( 5x10** n/cm');
Resulting low predicted shift in the reference nil-ductility temperature, RT,, (<60*F at 32 EFPY). is a technical report from General Electric Company (GE) that provides the details associated with the specific conditions that exist at LGS andLprovides the basis for a revised withdrawal schedule with the first capsule being removed at 15 EFPY. The withdrawal schedule for the second and
' third capsule remains unchanged.
The first surveillance capsule testing schedule should be developed to measure a significant portion of the fracture toughness change, as measured by ART,.r.
Since the limiting plate material for the LGS Unit I vessel has a low expected ART,, of 34*F (at 1/4T) over the life of the plant, the recommended schedule should be; designed to ceasure a majority of the ART,, of the plate material.
, Given the low expected sh'ift, a criteria of 75% of the expected shift in RT,,
was_ selected to determine the revised schedule.
For LGS Unit 1, 75% of the expected shift is 0.75(34) - 26'F.
Using a criteria of 75% of the expected shift (26'F), the capsule will experience this shift for the plate material at approximately 15 EFPY.
As per Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, Regulatory Position 2, " Surveillance Data Available,"
the collection of two 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 revi.se the Pressure-Temperature (P-T) curves (Technical Specification Figure 3.4.6.1-1).
Without two 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, Rev. 2,. Regulatory Position 1, " Surveillance Data Not Available." These methodologies use plant specific chemistry and fluence values to determine a calculated shift in RT,.r.
A " margin" term is then added to obtain conservative, upper-bound values of adjusted reference temperature.
The existing LGS Unit 1 P-T curves are based upon the Regulatory Position 1 methodology, and are currently valid up to 12 EFPY, With first capsule removal at eithet 10 nr 15 EFPY, the existing P-T curves will require a revisio6, prior to reaching 12 EFPY, based upon the calculational methodologies as contained in the Regulatory Guide 1.99, Rev. 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 the curves beyond 12 EFPY.
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Docket No. 50-352 4
Lic:nse No. NPF-39
.LER 87-032-00 evaluated / documented the: impact of no first cycle flux wire data on continued plant operation.
Part of the justification provided' relied on the analysis of-the flux wires within the first surveillance capsule.
Although-it-was assumed that the first surveillance specimen would be removed at 10 EFPY, the flux wires _ within the surveillance capsules, in conjunction with the daily power history, will be able to provide 'an accurate value for LGS Unit I vessel fluence at 15 EFPY.
As specified in Attachment 3, justification for the revision of the first capsule withdrawal' schedule from 10 EFPY to 15 EFPY is based on the following:
1.
Based on ART calculations performed in accordance with Regulatory Guide 1.99, Rev. 2, the shift (ART,, + margin) for the LGS Unit I vessel plate is calculated to be 68'F at 32 EFPY.
If the first capsule is removed at 10 EFPY, the actual shift (predicted to be 21*F) may not be large enough i
to be differentiated from the data scatter, since the predicted fluence on the capsule at 10 EFPY (4.9x10" n/cm' is low, and the chemistry of the LGS Unit I vessel plate material is excellent (0.01-0.12% copper).
Thus, the data obtained may not be useful for predicting the material behavior, as it may not be distinguishable from the unirradiated data.
Therefore, removal of the 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 RT,, shifts and measured RT,, shifts from other BWR surveillance capsules, the predicted shifts bound the measured results.
Figure 2-1 of Attachment 3 is a plot of actual shift measurements versus predicted shifts (calculated per Regulatory Guide 1.99, Rev. 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, Rev. 2 is based upon a conservative calculation, and is expected to bound the actual fluence.
4.
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, Rev. 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 surveillance requirement 4.4.6.1.3,. if required, and will meet the requirements of 10 CFR 50 Appendix H and ASTM E-185.
5.
The Supplemental Sur_veillance 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 at Cooper and Oyster Creek.
. SSP specimens will provide early test data for vessel plate, which is similar to the LGS Unit I surveillance plate. The SSP will suoplement the LGS Unit I surveillance program by providing timely detection of
-anomalous RT,, shifts,- should'any occur.
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Docket'No. 50-352 r
License No. NPF-39 The combination of the low expected shift for 1[he 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 1.
A revision to TS Surveillance Requirement 4.4.6.1.4 is also proposed. This 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 change will update-the TS Surveillance Requirements to reflect the elimination of the use of first cycle flux data, since this data could not be collected, and the addition of the use of flux data from the flux wires located within the surveillance capsulet (LER87-032-00),
Information Sucoortina a Findino of No Sianificant Hazards Consideration =
We have concluded that the proposed changes to the Limerick Generating Station-(LGS), Unit 1 Technical Specifications (TS) which will revise TS Table 4.4.6.1.3-1 to change the withdre.wal schedule for the first capsule from 10 Effective Full Power Years-(EFPY) to 15 EFPY, and TS Surveillance Regtirement 4.4.6.1.4 to remove reference to 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 10 CFR 50.92 is provided below:
1.
The proposed Technical Specifications (TS) chanaes do not involve a sianificant increase in the probability or conseauences of an accident oreviousiv 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 SAR. The changes revise the withdrawal schedule for the reactor vessel material surveillance capsules from 10 Effective Full Power Years (EFPY) to 15 EFPY. The capsules are not an initiator of any previously analyzed accident nor does the withdrawal schedule of the surveillance capsule affect the probability or consequences of any previously analyzed accident.
These changes will not affect the Pressure-Temperature (P-T) limits as given in LGS Technical Specification (TS) Figure 3.4.6.1-1 and 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 RT, as described in ASME Section III,
? Appendix G.
Changes in the fracture toughness properties of reactor pressure vessel (RPV) beltline materials, resulting from neutron L
-irradiation and the thermal environment, are monitored by a surveillance program in compliance with the requirements of 10 CFR 50 Appendix H.
The effect of neutron fluence on the shift in the RT, is predicted by methods'given in Regulatory Gtide 1.99, Rev. 2.
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Docket No. 50-352 o
License No. NPF-39 As detailed in Attachment 3, for LGS Unit 1, the combination of low expected RT, 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 10 EFPY to 15 EFPY--will result in a more credible set of surveillance data while ensuring the continued safe operation of LGS Unit 1.
LGS's current P-T limits were established based on adjusted reference temperatures developed in accordance with the procedures prescribed in Regulatory Guide 1.99, Rev. 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 IR06, 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 the capsule constitutes one set of credible surveillance data.
The curves will continue to be established in accordance with Regulatory Position 1 procedures.
As per Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, Regulatory Position 2, " Surveillance Data Available," the collection of two 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 Pressure-Temperature (P-T) curves (Technical Specification Figure 3.4.6.1-1).
Without two 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, Rev. 2, Regulatory Position 1, " Surveillance Data Not Available." These methodologies use plant specific chemistry and fluence values to determine a calculated shift in RT,.
A margin" term is then added to obtain conservative, upper-bound values of adjusted reference temperature.
The existing LGS Unit 1 P-T curves are currently valid up to 12 EFPY, With first capsule removal at either 10 gr 15 EFPY, the existing P-T curves will require a revision prior to reaching 12 EFPY based upon the calculational methodologies as contained in the Regulatory Guide 1.99, Rev. 2, Regulatory Position 1, " Surveillance Data Not Available."
Therefore, the 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 12 EFPY.
The fluence data as determined from the surveillance capsule flu wires at 15 EFPY will provide an accurate indication of neutron fluence.
In accordance with Regulatory Guide 1.99, Rev. 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 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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Docket No. 50-352 o
LicGnse No. NPF-39 These changes will not affect any plant safety limits or limiting conditions of operation.
The proposed changes will not affect reactor pressure vessel performance as they do not involve any physical changes, and LGS P-T limits will remain conservative in accordance with Reg.
Guide 1.99, Rev. 2 requirements. The proposed changes will not cause the RPV or interfacing systems to be operated outside of their design or testing limits.
The proposed changes do not increase the 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 RPV 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 Unit I reactor pressure vessel until the data from the first LGS Unit I capsule is available. The proposed changes provide the same level of confidence in the integrity of the vessel.
Therefore, the proposed TS changes do not involve an increase in the probability or consequences of an accident previously evaluated.
2.
The oroposed TS chanaes do not create the oossibility 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 will revise the withdrawal schedule for the first reactor pressure vessel (RPV) material surveillance capsule from 10 Effective full Power Years (EFPY) to 15 EFPY, These 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 changes.
No new accident types or failure modes will be introduced as a result of the proposed changes.
LGS's current Pressure-Temperature (P-T) limits were established based on adjusted reference temperatures developed in accordance with the procedures prescribed in Regulatory Guide 1.99, Rev. 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 IR06, 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 the capsule constitutes one set of credible surveillance data. The curves will continue to be I
established in accordance with Regulatory Position 1 procedures.
Tne existing LGS Unit 1 P-T curves are currently valid up to 12 EFPY.
With first capsule removal at either 10 gr 15 EFPY, the existing P-T curves will require a revision, prior to reaching 12 EFPY, based upon l
the calculational methodologies as contained in the Regulatory Guide i
j 1.99, Rev. 2, Regulatory Position 1, " Surveillance Data Not Available."
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i D:cket No. 50-352
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Lic nse No. NPF-39 s
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Therefore, the Technical Specification (TS) revision to the first capsule' withdraw 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 12 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, Rev. 2, Regulatory Position 1 methodology, data from thesa flux wires will permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS Surveillance Requirement 4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50 Appendix H and ASTM E-185.
The potential for reactor pressure vessel (RPV) failure will continue to be adequately assessed by the proposed withdrawal schedule. As detailed in Attachment 3, the combination of the low expected shift for the 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 1.
The proposed changes provide the same level of confidence in the integrity of the RPV.
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 chanaes do not involve a sianificant reduction in a marain of safety.
The proposed thanges to the TecMical Specifications (TS) do not reduce the margin of safety as defined n 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 canditions, system response time, or in any other parameter affecting the accident analyses supporting the Bases of any TS.
The propo' sed changes do not involve revision of the P-T limits but rather a revision of the withdrawal schedule for the first surveillance capsule. The current P-T limits were established based on the adjusted reference temperatures for vessel beltline materials calculated in accordance with Regulatory Position 1 of Reg. Guide 1.99, Rev. 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 or more credible surveillance data sets become available. When two or more credible surveillance data sets become available, P-T limits will be revised as prescribed by Regulatory Position 2 of Reg.-Guide 1.99, Rev. 2 or other NRC approved guidance.
h T e current P-T limit curves are inherently conservative and provide sufficient margin to ensure the integrity of the reactor pressure vessel. The proposed 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.
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faut?t No. 50-352 i
. De~ a No. NPF-39 In accordance with Regulatory Guide 1.99, Rev. 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 Surveillance Requirement
'.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 Suonortina an Environmental Assessment 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 1-TS conform to the criteria for " actions eligible for categorical exclusion" as specified in 10 CFR 51.22(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 l
section. The proposed ' changes do not involve a significant change in the i
types or 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 1 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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. ATTACHMENT 2 -
LIMERICK GENERATING STATION UNIT 1 DOCKET NO. 50-352 LICENSE NO. NPF-39 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 97-02-1 i
n LIST OF AFFECTED PAGES UNIT 1 3/4 4-19 3/4 4-21
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. s REACTOR CDOLANT SYSTEM SURVEILLANCE REQUIREPENTS (Continued) 4.4,6.1.2. The reactor coolant system temperature and pressure shall be deteminen to be to the right of the criticality limit line of Figure 3.4.6.1-1
. curve C within 15 minutes prior to the withdrawal of, control rods to.
l bring the reactor to criticality and at least once per 30 minutes during system heatup.
4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed cnd examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1.
The.results of these exas nn shall he used._to u:uinte.
the curves of Figure 3.4.6.1-1.
i.oento werd twa s waiticades) ^*E c.Aesvi.aL _ -
4.4.6.
4 The reactor flux wire specimens"sh01 be removed et th: f'r:t 7:f;;1'n, nd examined to determine reactor pressure vessel fluence as a function of time and power level and used to modify Figure B 3/4 4.6-b The results of these fluence determinations shall be used to adjust the curves of
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fD id "*ENME WNW NE tr#EDuM required.
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.'3 - g The reactor vessel flange and had flange temperature [ hall be veM 4.4.6.1.5 to be greater than or equal to 80*F:
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In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
1.
i 100*F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
190*F, at least once per,30 minutes.
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"hMn 50 minutes prior to and at least once per 30 minutes during tensioning of tne reactor vessel head bolting studs.
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ATTACHMENT 3 LIMERICK GENERATING STATION '
UNIT 1 i
. DOCKET NO. 50-352 LICENSE NO. NPF-39 TECHNICAL SPECIF! CATIONS CHANGE REQUEST NO. 97-02-1 GE Report GE-NE-B1100786-01R1
" SURVEILLANCE SPECIMEN PROGRAM EVALUATION FOR LIMERICK GENERATING STATION, UNIT 1,"
dated December 1997.
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