ML20154H966
| ML20154H966 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/20/1988 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20154H938 | List: |
| References | |
| NUDOCS 8805260129 | |
| Download: ML20154H966 (10) | |
Text
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' REACTOR COOLANT SYSTEM 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, cooldovn, criticality, and inservice leak and hydrostatic testing with
- a. A maximum heatup of 100*F in any one hour period.
- b. A maximum cooldown of 100*F in any one hour period.
- c. A maximum temperature change of less than or equal to 10*F in any one i
hour period during inservice hydrostatic and leak testing operations l above the heatup and cooldown limit curves.
i APPLICABILITY: At all times.
i ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation or inspection to determine the effects of the out-of-limit condition on the
, fracture toughness of the Reactor Pressure Vessel; determine that the Reactor l Pressure Vessel remains acceptable for continued operation or be in at least HOT l STANDBY vithin 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 foll8ving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS l
l 4.4.10.1.1 The Reactor Coolant System temperature 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.
i 4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.
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FARLEY-UNIT 1 3/4 4-27 AMENDMENT NO.
8805260129 880520 PDR ADOCK 05000348
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a TABLE 4.4-5 THIS PAGE HAS BEEN DELETED l
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t FARLEY-UNIT 1 3/4 4-20 AMENDMENT NO.
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REACTOR COOLANT SYSTEH BASES l
Values results of 6RT,$,
from t e determined in this manner material surveillance may be program, used until evaluated the nextto according ASTM E185-82, are available. Capsules vill be removed in accordance with the requirements of ASTH E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in FSAR Section 5.4. l The heatup and cooldown curves must be recalculated when the 6RT determinedfromthesurveillancecapsuleexceedsthecalculated1kk,,,
for the equivalent capsule radiation exposure.
Allovable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASHE Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in VCAP-7924-A.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic frseture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the vall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel vall as well as at the outside of the vessel vall. The dimensions of this postulated crack, referred to in Appendix G of ASHE Section III as the reference flav, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RT,,,, is used and this includes the radiation induced shift, 6RT , corresponding to the end of the period for which heatup and cooldo,,, vn curves are generated.
FARLEY-UNIT 1 B3/4 4-8 AMENDHENT NO.
REACTOR COOLANT SYSTEM 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 maximum heatup of 100*F in a'.y one hour period,
- b. A maximum cooldovn of 100*F in any one hour period.
- c.
- A maximum 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 cooldovn limit curves.
APPLICABILITY: At all times.
ACTION:
Vith any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation or inspection to determine the effcets of the out-of-limit 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 yithin 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, vitbin the foll8ving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System temperature 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.
4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.
FARLEY-UNIT 2 3/4 4-27 AMENDMENT NO.
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TABLE 4.4-5 THIS PAGE HAS BEEN DELETED FARLE/-UNIT 2 3/4 4-28 AMENDMENT NO.
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REACTOR COOLANT SYSTEM BASES
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g determined in this manner may be used until the next Valuesof6RT,$ematerialsurveillanceprogram,evaluatedaccordingto results from t ASTM E185-82, are available. Capsules vill be removed in accordance with l the requirements of ASTM E185-82 end 10 CFR 50, Appendix H. The surveillance specimen withdraval schedule is shown in FSAR Section 5.4. l The heatup and cooldown curves must be recalculated when the 6RT determinedfromthenextsurveillancecapsuleexceedsthecalculaied 6RT,3, for the equivalent capsule radiation exposure.
Allovable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code a: required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in VCAP-7924-A.
The general method for calculating heatup and cooldovn limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) i technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the vall thickness, T, and a length of 3/2T is a'eumed to exist at the inside of the vessel vall as well as at the outside of the vessel vall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flav, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RT,gg, is used and this includes the radiation induced shift, 6RT , corresponding to the end of the period for which heatup and cooldo,n,, v curves are generated.
FARLEY-UNIT 2 B3/4 4-8 AMENDMENT NO.
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a ATTACllMFRr 2
Significant Hazards Evaluation Pursuant to 10 CFR 50.92 for the Deletion of the Reactor Vessel Material Surveillance Specimen Vithdrawal 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 the portion of Surveillance Requirement 4.4.10.1.2 which references Table 4.4-5 and Figures 3.4-2 and 3.4-3, and deletion of Table 4.4-5 itself. In addition, the corresponding Bases is revised to eliminate the reference to Table 4.4-5 and indicate that the information previously provided in the table vill 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 CFR 50 Appendix H and has been reviewed and approved by the Office of Nuclear Reactor Regulation. The schedule for removal of reactor vessel surveillance specimens is contained 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 information contained in Table 4.4-5.
For example, since surveillance specimens are removed at the refueling outage nearest the scheduled removal exposure, the actual exposure at removal vill likely vary from that indicated in the schedule. Following removal of each specimen, the schedule for withdraval of remaining specimens is reviewed to ensure that the requirements of 10 CFR Appendix H are satisfied. Updating the surveillance specimen withdrawal schedule to reflect the actual time of specimen removal currently requires a license amendment.
Deletion of Table 4.4-5 from the Technical Specifications vill allow future adjustments to the withdrawal schedule, including the lead factors, to be made without submittal of a license amendment request. It is anticipated that future changes to the surveillance specimen withdraval schedule vill only be necessary as a result of the analysis of surveillance specimens. Since the Code of Federal t Regulations requires that the results of each surveillance specimen analysis be submitted to the NRC, the reactor vessel material surveillance program information vill continue to be provided to the NRC. It should be noted that the Technical Specification Bases vill 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 information currently included in Table 4.4-5 vill be added to the FSAR. Removal of this information from the Technical Specifications vill obviate the unnecessary use of licensee and NRC resources to process future licer.se amendments. In addition, deletion of this material vill enhance the usability of the Technical Specifications by plant operators resulting in an incremental benefit to plant safety.
Surveillance Requirement 4.4.10.1.2 requires that surveillance specimens be removed 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 system (RCS) heatup and cooldovn limitation
B-Significtnt H;2crds Evolustion Pursuant to 10 CFR 50.92 for ths Deletion of the Reactor Vessel Material Surveillance Specimen Vithdraval Schedule from the Technical Specifications Page 2 curves in Technical Specifications (Figures 3.4-2 and 3.4-3). All of the conditions of this Surveillance Requirement are inherent in the Code of Federal Regulations. The schedular requirements for withdraval of specimens are included in ASTM E 185 which is referenced in Appendix H. Rules for the application of the results of material examinations used in the determination of heatup and cooldown limitations are found in 10 CFR 50 Appendix G vhich is also 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 must continue to be reviewed as results from the material surveillance program are obtained. Thus, the conditions of Surveillance Requirement 4.4.10.1.2 are redunde.nt to the Code of Federal Regulations.
It is anticipated that NRC approval of this requested Technical Specification change vill occur subsequent to Revision 6 of the Farley Nuclear Plant FSAR Update (July 1988). Revision 6 vill add the information currently included in 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 withdraval of surveillance specimens is shown in FSAR Section 5.4.
The Bases vill retain the reference to 10 CFR 50 Appendix H and ASTM E 185-82.
It should be noted that two minor editorial changes are being made on B 3/4 4-8.
Specifically, the word "next" is being added as the last vord 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 applict.ble version of ASTM E 185 is the 1932 edition. These changes are strictly editorial and are requested to restore the similarity of the Unit 1 and Unit 2 Technical Specifications.
Analysis Alabama Power Company has revieved the requirements of 10 CFR 50.92 as they relate tonot change this proposed Technical Specification change and considers the proposed to involve a significant hazards consideration. In support of this conclusion the following analysis is provided:
- 1) The proposed change does not significantly increase the probability or consequences of an accident previously evaluated because the reactor vessel material surveillance program is not affected by this proposed change. Implementation of the proposed change vill delete a license requirement that is redundant to the Code of Federal Regulations. Thus, this proposed Technical Specification is considered to be administrative in nature.
- 2) The proposed change vill not create the possibility of a new or different kind of accident frcn any accident previously evaluated because implementation of this change vill not alter plant configuration or mode of operation. Compliance with existing regulations vill ensure continued confidence in reactor vessel material properties.
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Significtnt HIzsrds Evaluation Pursuant to 10 CFR 50.92 for ths
' Deletion of the Reactor Vessel Material Surveillance Specimen Vithdrawal Schedule from the Technical Specifications Page 3
- 3) The proposed change vill not involve a significant reduction in the margin of safety because the evaluation of reactor vessel material embrittlement is not altered 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 vill actually enhance the usability of the Technical '
Specifications by plant operators resulting in an incremental benefit to plant safety.
Conclusion .
Based upon the analysis provided herewith, Alabama Power Company has determined that the proposed Technical Specification change vill not significantly increase the probability or consequences of an accident previously 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 does not involve a significant hazards consideration.
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