ML20148T448

Proposed Revs to Tech Specs,Deleting Surveillance Requirement 4.4.10.1.2 & Table 4.4-5 Re Surveillance Specimen Withdrawal Times.Related Info Encl

ML20148T448
Person / Time
Site:	Farley
Issue date:	01/28/1988
From:	ALABAMA POWER CO.
To:
Shared Package
ML20148S985	List: ML20148T448
References
NUDOCS 8802030309
Download: ML20148T448 (10)

Text

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REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMIT!NG CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be 1.imited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A naximum heatup of 100'F in any one hour period,

b. A naximun cooldown of 100'F in any one hour period,

c. A naximum temperature change of less than or equal to 10'F in any one hour period during inservice hydrostatic and leak testing operations I above the heatup and cocidown limit curves.

APPLICABILITY: At all times.

1 ACTION:

l With any of the above limits exceeded, restore the temperature and/or pressure I to within the limit within 30 minutes; perform an engineering evaluation or

! inspection to determine the effects of the out-ofalimit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200*F and 500 psig, respectively, within the fol5! wing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4 SURVE!LLANCE REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System terperature and pressure shall be determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

FARLEY-UNIT 1 3/4 4-27 AMENDMENT NO.

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O. 9 TABLE 4.4-5 THIS PAGE HAS BEEN DELETED I

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FARLEY-UNIT 1 3/4 4-28 AMENDMENT NO.

REACTOR COOL, ANT SYSTEM BASES Values of LRTndt determined in this manner may be used until the next results from the naterial surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50 Appendix H. The surveillance specinen withdrawal schedule is shown in FSAR Section 5.4.

The heatup and cooldown curves must be recalculated when the LRTndt deternined from the surveillance capsule exceeds the calculated ARTndt for the equivalent capsule radiation exposure.

) Allowable pressure-tenperature relationships for various heatup and l cooldown rates are calculated using nethods derived from Appendix G in Section !!! of the ASME Boiler and Pressure Yessel Code as required by Appendix G to 10 CFR 50 and these rethods are discussed in detail in WCAP-7924-A.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a sent-elliptical surface defect with a depth of one-quarter of the wall thickness. T and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dirensions of this postulated crack, referred to in Appendix G of ASt'E Section !!! as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide suf ficient safety nargins for protection against non-ductile failure. To assure that the radiation embrittlerent ef fects are accounted for la the calculation of the limit curves, the most limiting value of the nil ductility reference tenperatu e, RTndt, is used and this includes the radiation induced shif t, tRindt corresponding to the end of the period for which heatup and cooldown curves are generated.

FARLEY-UNIT 1 B3/4 4-8 AMEN 0 MENT NO.

REACTOR COOLANT SYSVEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A naximum heatup of 100'F in any one hour period,

b. A naximum cooldown of 100*F in any one hour period,

c. A raximun temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.  :

i ACTION: ,

With any of the above limits exceeded, restore the tenperature and/or pressure  :

to within the limit within 30 minutes; perform an engineering evaluation or l inspection to determine the effects of the cut-of-limit condition on the  !

fracture toughness of the Reactor Pressure Vessel; determine that the Reactor 1 Pressure Vessel remains acceptable for continued operation or be in at least HOT .

STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less r than 200*F and 500 psig, respectively, within the fol5$ wing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. l

$URVE!LLANCE REQUIREMENTS  !

........ ..................a........... ................. .. .. . ..... ,

4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be f determined to be within the limits at least once per hour during system heatup, l cooldawn, and inservice leak and hydrostatic testing operations. ,

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FARLEY-UNIT 2 3/4 4-28 AMENDMENT No.

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REACTOR COOLANT SYSTEM BASES

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Values of ARTndt determined in this manner may be used until the next results from the raterial surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirenents of ASTM E185-82 and 10 CFR 50 Appendix H. The surveillance specimen withdrawal schedule is shown in FSAR Section 5.4 The beatup and cooldown curves must be recalculated when the ARindt determined from the next surveillance capsule exceeds the calculated LRTndt for the equivalent capsule radiation exposure.

Allowable pressure-terperature relationships for various heatup and cooldown rates are calculated using nethods derived from Appendix G in Section !!! of the ASPE Boiler and Pressure Vessel Coce as required by Appendix G to 10 CFR 50 and these nethods are discussed in detail in WCAP-7924-A.

The general nethod for calculating heatup and cooldown linit curves is based upon the principles of the linear elastic fracture nechanics (LEFM) technology. In the calculation procedures a semi-elliptical surf ace defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assuned to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dinensions of this postulated crack, referred to in Appendix G of ASME Section !!! as the reference flaw, arply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provice sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlerent ef fects are accounted for in the calculation of the limit curves, the nost limiting value of the nil ductility reference terperature. RTndt, is used and this includes the radiation induced shift, aRTndte corresponding to the end of the period for which heatup and cooldown curves are generated.

FARLEY-UNIT 2 83/4 4-8 AMENDMENT NO.

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Significant Hazards ' Evaluation Pursuant to 10 CFR 50.92 for the Deletion of the Reactor Vessel Material Surveillance Specimen Withdrawal Schedule from the Technical Specifications Proposed Change The purpose of this proposed change is to delete the reactor vessel surveillance specimen withdrawal schedule from the Technical Specifications.

This change involves the deletion of Surveillance Requirenent 4.4.10.1.2 and Tablo 4.4-5. In addition, the corresponding Bases is revised to eliminate the reference to Table 4.4-5 and indicate that the inforration previously provided in the table will be added to the FSAR.

Background

The Farley Nuclear Plant Unit 1 and 2 program for surveillance of reactor vessel materials is governed by 10 CFP 50 Appendix H and has been reviewed and approved by the Office of Nuclear Reactor Regulation. The schedule for renoval of reactor vessel surveillance specinens is conuined in Technical Specification Table 4.4-5 and complies with the guidance of ASTM E 185 as directed by 10 CFR 50 Appendix H. Periodically the need arises to update the informtion contained in Table 4.4-5. For example, since surveillance spectrens are reuved at the refueling outage nearest the sr.hedulrd removal exposure, the actual exposure at removal will likely vary from that in *icated in the schedule. Following renoval of each specimen, the schedule for withdrawal of remaining specinens is reviewed to ensure that the requirenents of 10 CFR Appendix H are satisfied. Updating the surveillance specimen withdrawal schedule to reflect the actual time of spectren removal currently requires a Itcense arendnent.

Deletion of Table 4.4-5 from the Technical Specifications will allcw future adjustnents to the withdrawal schedule, including the lead factors, to be m de without submittal of a license arendrent request. It is anticipated that future changes to the surveillance spectren withdrawal schedule will only be necessary as a result of the analysis of surveillance spectrens. Since the Code of Federal Regulations requires that the results of each surveillance spectren analysis be submitted to the ARC. the reactor vessel material survetilance progran inforr.ation will continue to be provided to the NRC. It should be noted that the Technical Specification Bases will retain the description of the reactor vessel material surveillance program including the reference of 10 CFR 50 Appendix H and ASTM E 185-82. The informtion currently included in Table 4.4-5 will be added to the FSAR. Rer. oval of this inform tion from the Technical Specifications will obviate the unnecessary use of licensee and ARC resources to process future license arendrents. In addition, deletion of this raterial will enhance the useability of the Technical Specifications by plant operators resulting in an increcental benefit to plant safety.

Surveillance Requirenent 4.4.10.1.2 requires that surveillance specinens be renoved in accordance with the schedule in Table 4.4-5, examined in accordance with 10 CFR 50 Appendix H and that the results of the capsule examinations be used to update the reactor coolant systen (RCS) heatup and cooldown limitation I

Significant Hazard Evaluation Pursuant to 10 CFR 50.92 for the i Deletion of the Reactor Vessel Material Surveillance Specimen Withdrawal Schedule from the Technical Specifications l >

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curves in Technical Specifications (Figures 3.4 2 and 3.4-3). All of the I i

conditions of this Surveillance Requirenent are inherent in the Code of Federal l

. Regulations. The schedular requirements for witharcwal of specinens are i l included in ASTM E 185 which is referenced in Appendix H. Rules for the  :

l application of the results of material examinations used in the determination of heatup and cooldown Ifmitations are found in 10 CrR 50 Appendix G which is also ,

i referenced by 10 CFR 50 Appendix H. Since Appendix G specifies the pressure and '

temperature limits for the reactor vessel based on the material properties, the Technical Specification heatup and cooldown curves rust continue to be reviewed as results from the naterial surveillance program are obtained. Thus, the l conditions of Surveillance Requirerent 4.4.10.1.2 are redundant to the Code of i federal Regulations.  !

It is anticipated that NRC approval of this requested Technical Specification I change will occur subsequent to Revision 6 of the Farley Nuclear Plant FSAR

. Update (July 1988). Revisio9 6 will add the information currently included in l Technical Specification Table 4.4-5 to Section 5.4 of the FSAR. Accordingly,  !'

the proposed change to Technical Specification Bases 3/4.4.10 indicates that the schedule for withdrawal of surveillance spectrens is shown in FSAR Section 5.4.

The Bases will retain the reference to 10 CFR 50 Appendix H and ASTM E 185-82. j It should be noted that two minor editorial changes are being rade on B 3/4 I 4-8. Specifically, the word "next" is being added as the last word on the first line of Unit 1 page B 3/4 4-8 The first sentence of Unit 2 page B 3/4 4-8 is being revised to indicate that the applicable version of ASTM E 185 is t5e 1982 edition. These changes are strictly editorial and are requested to restore the ,

siellarity of the Unit 1 and Unit 2 Technical Specifications. t l

I Analysis f

Alabama Power Corpany has reviewed the requirenents of 10 CFR 50.92 as they relate to this proposed Technical Specification change and considers the proposed change not to involve a significant hazards consideration. In support of this conclusion the following analysis is provided:

• ) The proposed change does not significantly increase the probability or consequences of an accident previously evaluated because the reactor vessel raterial surveillance progran is not af fected by this proposed change. Implerentation of the proposed change will delete a license requirerent that is redundant to the Code of Federal Regulations. Thus, this proposed Technical Specification is considered to be administrative in nature.

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Significant Hazards Evaluation Pursuant to 10 CFR 50.92 for tha Deletion of the Reactor Yessel Material Surveillance Specimen Withdrawal Schedule from the Technical Specifications Page 3

2) The proposed change will not create the possibility of a new or dif ferent kind of accident frca any accident previously evaluated because implementation of this shange will not alter plant configuration or mode of operation. Corapliance with es,isting regulations will ensure continued confidence in reactor vessel material properties.

3) The proposed change will not involve a significant reduction in the margin of safety because the evaluation of reactor vessel material embrittlenent is not a!tered by this change. Additionally, Surveillance Requirement 4.4.10.1.2 and Table 4.4-5 are not beneficial to the primary user of the Technical Specifications (i.e., the reactor operator). Thus, deletion of this material will actually enhance the useability of the Technical Specifications by plant op9rators resulting in an incremental benefit to plant safety.

Conclusion *-

Based upon the analysis prow'fded herewith, Alaban.1 P' ower Corpany has determined that the proposed Technical SpeU4fication change will not significantly increase the probability or consequences of an accident prevously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety. Therefore, Alabama Power Company has determined that the proposed change' meets the requirements of 10 CFR 50.92 and dces not involve a significant hazards consideration.

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