ML17312B692
| ML17312B692 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/11/1997 |
| From: | Thomas K NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17312B693 | List: |
| References | |
| NUDOCS 9710010244 | |
| Download: ML17312B692 (36) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.
DOCKET NO.
STN 50-528 PALO VERDE NUCLEAR GENERATING STATION UNI'T NO. I AMENDMENT TO FACILITY OPERATING 'LICENSE Amendment No. ii3 License No.
NPF-41 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District. El Paso Electric Company.
Southern Ca1ifornia Edison
- Company, Public Service Company of New Mexico.
Los Angeles Department of Water and Power.
and Southern California Public Power Authority dated Decembei 27.
- 1996, as supplemented by letter dated August 22,
- 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I:
2.
B.
The facility will operate in conformity with the application. the provisions of the Act, and the rules and regulations of. the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public.
and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of'his amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No.
NPF-41 is hereby amended to read as follows:
97i00i0244 9709ii PDR ADQCK 05000S28 P
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 113, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license.
APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection. Plan, except.where otherwise stated in specific license conditions.
3.
This license amendment is effective as of the date of issuance
=to be implemented within 30 days from,its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION M.t~~
Kristine M. Thomas, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of'uclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 11, 1997
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
113 TO FACILITY OPERATING LICENSE NO.
STN 50-528'eplace the following pages of'the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identifi'ed by Amendment number and contain marginal li.nes indicating the areas of change..
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE 3/4 6-4 B 3/4 6-2 6-25 INSERT 3/4 6-4 B 3/4 6-2 6-25
0 Qi h
I
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE,with a.
Both doors closed except when the air lock is being used for normal transit entry and exit through the containment.
then at least one air lock door shall be closed, and b.
An overall air lock leakage.rate of less than or equal to 0.05 L, at P,,
52 psig.
APPLICABILITY:
MODES 1, 2, 3, and.4.
ACTION:
a.
With one containment air lock door inoperable:
1.
Maintain at least the OPERABLE air lock door closed* and either restore the inoperable.air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the.OPERABLE air lock door closed.
Operation may then continue until.performance of the next required overall air lock leakage test provided that the OPERABLE.air lock. door is verified to be locked closed at least once per 31 days, or 2.
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 in COLD SHUTDOWN within the foll'owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.
The provisions of'pecification 3.0.4-are not applicable.
b.
With.the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE UIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
In accordance with the Containment Leakage Rate Testing Program at periodic intervals and following each closing as specified.
The provisions of Specification 4.0.2 are not applicable to the overal.l air lock leakage test.
- Except during entry to repair an inoperable inner door
. for a cumulative time not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per year.
PALO VERDE
-, UNIT 1 3/4 6-4 Amendment No. 47-.%8.ii~
0
ASES Primary CONTAINHENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
The limitations on containment leakage rates ensure that the total'ontainment leakage volume will'ot exceed the value assumed in the safety analyses at the peak accident pressure, Pa.
As an added conservatism,, the measured overall integrated leakage rate is further limited to.less than or equal to 0.75 L or less than or equal to 0.75 L as applicable during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Option B of. Appendix J, of 10 CFR Part. 50 and Regulatory Guide 1.163, dated S'eptember 1995.
0 OC The limitations on closure and leak rate for the containment:air locks are required to meet the restrictions on CONTAINHENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provides assurance that
'the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
PALO'VERDE - UNIT I B 3/4 6-1 Amendment No;103
CONTAINMENT SYSTEMS BASES 3/4 6 1 4 INTERNAL PRESSURE The limitations ion containment internal pressure ensure that (1) the containment structure is prevented from exceeding its design negative pressure differential with respect, to the outside at;mosphere of 4 psig and (2) the'ontainment peak pressure does not exceed the design pressure of 60 psig during LOCA conditions.
The maximum pressure increase expected to be obta.ined from a LOCA event is 49.5 psig.
The limit of'.5 psig for initial positive containment pressure will limit the total pressure to 52 p ig which is less than the design prlesdurh (60 psig) and is consistent with the safety analyses.
3/4.6. 1. 5 AIR TEiMPERATURE The limitation on containment average air temperature ensures that the overall containment average air temperature does, not exceed the initial temperature condition assumed.in the afety analysis.
3/4. 6.1.6 CONTAINMENT STRUCTUIRAI INTEGRITY This limitation ensures that the structural integrity of the conttainment~
will be maintained comIparable to the original design standards for the life of the facility.
Structural integr.ity is required to ensure that the conttainment~
will withstand the maximum pressure oi 52 psig in the event of a LOCA.
Ilhe containment design presslire is 60 psig.
Tf'ie measurement of containment tendon lift-off'orce; the tensile tests of the tendon wires or strands; the examination and test:ing of 'the sheathing filler grease; and the visual examination of tendon anchorage assembly
- hardware, surrounding concrete and the exterior surfaces of tlhe containment are sufficien<i: to demonstrate this capability.
The tendon wire or strand samples wil 1 also be ubjected to tests.
All of the required testing and visual examinations should be. performed in a time frame that permits a comparison of the results for the arne operating history.
The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with tlhe recommendations of ReghlaItoky Guide 1.35, "Inser vice Surveillance oi. Ungrouted Tendlons in Prestressed Concrete Containment; Structures."'evision l. 1974.
The requi red Special Report:s from any engineering evaluation of containment abnormalities shall include a description of the tendon condition,'he condition of the concrete (especially at tendon anchorages),
the inspection procedures.
the tolerances on cracking, the results of the engineering evaluation, and the corrective action 'aken.
PALO VERDE - UNIT 1 B 3/4 6-2 Amendment No. K,ii3
ADMINISTRATIVE CONTROLS.
2)
Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 3)
A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems; 4)
An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5)
An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 6)
A comparison of the predicted releases of radioactive materials.
in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to. when the changes are to be made; and 7)
An estimate of the exposure to plant operating personnel as a
result of the change.
- 6. 16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of containment as required by 10 CFR Part 50.54(o) and 10 CFR Part 50. Appendix J. Option B. as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1; 163.
"Performance-Based Containment Leak-Test Program,"
dated September, 1995 and ANSI/ANS-56. 8;1994.
The peak calculated. containment internal pressure for the design basis loss of coolant accident, P,. is 52 psig.
The maximum allowable containment leakage rate, L,, at P,. shall be 0.1X of containment air weight per day.
Leakage rate acceptance criteria are:
a.
Containment leakage rate acceptance criterion is ~ 1.0 L.
During the first unit.startup following testing in accordance with this rogram, the leakage rate acceptance criteria are ~ 0.6 L, for the ype B and Type C tests and ~ 0.75 L, for Type A tests:
b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is ~ 0.05 L, when tested at ~ P,,
2)
For each door, leakage rate is ~ 0.01 L, when pressurized to
> 14.5 + 0.5 psig.
PALO VERDE - UNIT 1 6-25 Amendment.No. ~2-,493,113
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.
DOCKET NO.
STN 50-529 PALO VERDE NUCLEAR GENERATING STATION UNIT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 106 License No.
NPF-51 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
B.
C.
D.
E.
The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself. and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison
- Company, Public Service Company of New Mexico. Los Angeles Department of Water and
- Power, and Southern California Public Power Authority dated December 27,
- 1996, as supplemented by letter dated August 22.
- 1997, complies with the standards and requirements of the Atomic
'nergy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application. the provisions of the Act. and the rules and regulations of'he Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public.
and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public:
and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No.
NPF-51 is hereby amended to read as follows:
0 Ql
,I
3.
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifi'cations contained in Appendix A, as revised through Amendment No.
- 106, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license.
APS shall operate the faci.lity in. accordance with the Technical Specifications and-the Environmental Protection Plan, except where otherwise stated in specific license conditions.
This license amendment is effective as of the date of issuance to be implemented within 30 days from its date of issuance.
FOR THE NUCLEAR REGULATORY.COMMISSION Kristine M. Thomas, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Speci'fications Date of Issuance:
September 11, 1997
el
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. io~
TO FACILITY OPERATING LICENSE 'NO. 'NPF-51 DOCKET NO.
STN 50-529 Replace the following pages of the Appendix A Technical, Specifi'cations. with the enclosed pages.
The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE 3/4 6-4' 3/4 6-2 6-25 INSERT 3/4 6-4 B 3/4 6-'2 6-25
41 0
A
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6. 1.3 Each containment air lock shall be OPERABLE with:
a.
Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.
An overall air lock leakage rate of less than or equal to 0.05 L, at P,,
52 psig.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
a.
With one containment air lock door inoperable:
1.
Maintain at least the OPERABLE air lock door closed* and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock door closed.
Operation may then continue until performance of the next requi red overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days. or 2.
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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3..
The provisions of Specification 3.0.4 are not applicable.
b.
With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE UIREMENTS 4.6. 1.3 Each containment air lock shall be demonstrated OPERABL'E:
a.
In accordance with the Containment Leakage Rate Testing Program at periodic intervals and following each closing as specified.
The provisions of Specification 4.0.2 are not applicable to the overall air lock leakage tests.
- Except during entry to repair an inoperable inner door, for a cumulative time not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per year:
PALO VERDE - UNIT 2 3/4 6-4 Amendment No. &2, io~
i
~
ASES Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part.100 during accident conditions.
NMENT The limitatipns on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P.
As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L or less than or equal to 0.75 L as applicable during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.-
The surveillance testing for measuring leakage rates are consistent with the requirements of Option B of Appendix J of 10 CFR Part 50 and Regulatory Guide 1.163, dated September 1995.
CO M NT The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of.the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
PALO VERDE - VNIT 2 B 3/4 6-1 Amendment No. 92
CONTAINMENT SYSTEMS 0 BASES 3/4.6. 1.4 INTERNAL PRESSURE':
The limitations on containment internal i)ressure ensure that (1) the containment structure
.is prevented frc~m exceedihg its'esign negative pressure differential with respect to the outside atmosphere of 4 psig and (2) the containment peak pressure does not exceed the design pressure of 60 psig during LOCA conditions,.
The maximum pressure increase expected to be obtained fr'om a LOCA event i<
49.5 psig.
The limit of 2.5 psig for initial pdsitive containment pressure will limit the total pressure to 52 psig which is less than the design pressure (60 psig) and is consistent with the safety anal~yses.'/4.6.1.5 AIR TElhPERAIIURE The limitation on containment average aid temperature ensures that the overall containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis.
3/4.6.1..6 CONTAINMENT STRUCTUIRAL IIITE'GRITS This limitation ensures that the structural integri'ty'f the containment will be maintained comparable to the original desi'gn standards For the life of the facility.
Structur al integr ity is requir0d to 'enSur'e that the containment will withstand the maximum pressure of 52 psig in t:he event of a LOCA.
The containment design pressure is 60 psig.
The measurement of containment tendon'ift-off force; the tensile tests of the te'nden wires or strands; the examination and testing of the sheathing fi'lier grease;
'and the visual examination of tendon
<anchorage assembly ha'rdware, surrounding concrete and the exterior surfaces of the containment are sufficient to demonstrate this capability.
The tendon wire or strand samples will also be subjected to tests.
All of'he requi red testing and visual examinations should be performed ih a time frame that permits a comparison of the'esults for the same operating history.
The Surveillance Requirements For demonptrating the containment's structural integrity are in compliance with t4e i"ec'ommendations of Regulatory Guide 1.35. "Inservice Surveillance of Ungrouted'endons in Prestressed Concrete Containment Structures,"
Revision 1, 1974.
The requi red Special Reports from any engineering evaluation of containment abnormalities shall. include a desc:ription of the tendon condition, the condition of the concrete (especially at tendon anchorages),
the inspection procedures, the tolerances cin cracking, the results of the engiraeering evaluation.
and the corrective actions taken.
PALO VERDE - UNIT 2 B 3/4 6-2 Amendment No.106
ADMINISTRATIVE CONTROLS 1)
A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.
2)
Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 3)
A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems; 4)
An evaluation of the change, which shows the predicted releases of radioactive materials in l,iquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5)
An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 6)
A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; and 7)
An estimate of the exposure to plant operating personnel as a
result of the change.
6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A.program shall be established to implement the l'eakage rate testing of
. containment as required by 10 CFR Part 50.54(o) and 10 CFR Part 50. Appendix J.
Option 8, as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1. 163.
"Performance-Based Containment Leak-Test Program," dated September, 1995 and ANSI/ANS-56.8-1994.
The peak calculated containment internal pressure for the design basis loss of coolant accident.
P,, is 52 psig.
The maximum allowable containment leakage rqte L,, at P,, shall be 0. 1X of containment air weight per day.
Leakage rate acceptance criteria are:
a..
Containment leakage rate acceptance criterion is ~ 1.0 L,.
During the first unit startup following testing in accordance with this rogram. the leakage rate acceptance criteria are ~ 0.6 L, for the ype 8 and Type C tests and ~ 0.75 L, f'r Type A tests; b.
Air lock testing. acceptance criteria are:
1)
Overall air lock leakage rate is ~ 0.05 L, when tested at ~ P,,
PALO VERDE - UNIT 2 6-25 Amendment No. 48-;92,106
~ggZ ggJ EgZgg REWED <Cyntj<nued')
2)
For each door, leakage rate 4s g 0.0l L, Ken pressurized:te h I4.5 k 0.5 [is)g.
'PALO VERDE -- UNIT 2 6-26 Amendment No.92
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 RIZONA. PUBLIC SERVICE COMPANY ET AL.
DOCKET NO.
STN 50-530 PALO VERDE NUCLEAR GENERATING STATION UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE 2.
Amendment No.
85 License No. NPF-74 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and
- Power, and Southern California Public Power Authority dated Decembei 27,
- 1996, as supplemented by letter dated August 22.
1997.
complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application, the provisions of the Act. and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public. and (ii) that such activities will be conducted in compliance with the Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common defense and security or to,the health and safety of the public:
and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No.
NPF-74 is hereby amended to read as follows:
I,
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A. as revised through Amendment No.
85
. and the Environmental Protection Plan contained in Appendix B, are, hereby incorporated into this license.
APS shall operate the facility,in accordance with the Technical Specifications and.the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective. as of the date of issuance to be implemented within 30 days from its date of issuance.
'OR THE NUCLEAR REGULATORY COMMISSION Kristine M. Thomas, Project Manager Project Directorate IV-2 Di'vision of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 11, 1997
Cl
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
85 TO FACILITY OPERATING LICENSE NO.
NPF-74 DOCKET NO.
STN 50'-530 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE 3/4 6-4 B 3/4 6-2 6-25 INSERT 3/4 6-4 B 3/4 6-2
'-25
0
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:
Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.
An overall air lock leakage rate of less than or equal to 0.05 L, at P,,
52 psig.
APPLICABILITY:
MODES 1, 2, 3, and.4.
ACTION:
a.
With one containment air lock door inoperable:
1.
Maintain at least the OPERABLE air lock door. closed* and either restore the inoperable.air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock-door closed.
Operation may then continue until performance of the next requi red overall air lock leakage test provided that the OPERABLE air lock. door is verified to be locked closed at least.once per 31 days, or 2'.
Be in at least HOT STANDBY within the next 6,hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.
The provisions of Specification 3.0.4 are not applicable.
b.
With the containment air lock inoperable, except as the result of an inoperable air lock door maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE UIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
In accordance with the Containment Leakage Rate Testing Program at periodic intervals and following each closing as specified.
'The provisions of Specification 4.0.2 are not applicable to the overall air lock leakage tests.
- Except during entry to repair an inoperable inner door, for a cumulative time not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per year.
PALO VERDE - UNIT 3 3/4 6-4 Amendment No. 74.85
O.
~ '
AS S
Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the.safety analyses.
This restriction, in conjunction with the leakage rate'imitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure,
'P.
As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L or less than or equal to 0.75 L as applicable during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Option B of Appendix J of 10 CFR Part 50 and Regulatory Guide 1.163, dated September 1995.
4.6.
N 0
The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
PALO VERDE - UNIT 3 B 3/4 6-1 Amendment No. 75
CONTAINMENT SYSTEMS 4l BASES 3/4.6. 1.4 INTERNAL PRESSURI=
The limitations on containment internal pressure ensure that (1) ithe containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 4 psig and (2) the containment peak pressure does not exceed the design pressure of 60 psiig during LOCA conditions.
The maximum,pressure increase expecte(i to be obtained from a LOCA~event; is I
49.5 psig.
The limit of 2.5 psig tor initial pcisi tive conta.inment pre.sure will limit the total pressure to 52 p.ig which is les,s than the design pressure (60 psig) and is consistent with the safety analyses.
3/4.6. 1. 5 AIR TEIMPERATURE The limitation on containment average air temperature ensure that the overall containment averaige air temperature does not exceed the initial temperature
.condi tion assumed in the. afety analysis.
3/4. 6. 1. 6 CONTAIINME:NT STRUCTURAI I NTE:GRllR'his limitation ensures that the structural integrity of'he containment will be maintained comparab'le to the original design standards for the life of the facility.
Structuiral integrity is required to eAsuire that the containment will withstand the maximum pressure of 52 psig in the event of a LOCA.
~ The
~
containment design pressure is 60 psig.
Tfie irneasurement of containment, tendon lift-off force; the tensile tests of the tendon wiires or strands; the
~
examination and testing of the sheathing filler grease; and the visual
~
examination of tendon anchorage assembly hardware.
surrounding concrete and the exterior surfaces of the containment are sufficient to demion. trate this capability.
The tendon wire or strand samples will ailso be subjected to tests.
All of the required testing and visuall examinations should be performed in a time frame that permits ai comparison of the results f'r the.arne operating
~
histor y.
The Surveillance Requi.rements for demotlstraiting the containment's structural integrity aire in compliance with the recommendations ot Reg&lattotty Guide 1.35.
"Inser vice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures,"
Revision 1.
1974.
The required Special Reports from any engineering evaluation of containment abnormalities, shall iinclude a description oF the tendon condition.
the condition of the concrete (especially at tendon anchorages),
the inspectio'n procedures, the tolerances on cracking, the results of the engineering~
evaluation, and the coirrective actions taken.
PALO VERDE - UNIT 3 B 3/4 6-2 Amendment No. 85
ADMINISTRATIVE CONTROLS 1)
A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.
2)
Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental info'rmation; 3)
A detai,led description of the equipment, components, and processes involved and the interfaces with other plant systems; 4)
An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5)
An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 6) 7)
A comparison of the predicted releases of radioactive materials.
in liquid and gaseous etfluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; and An estimate of the exposure to plant operating personnel as a
result of the change.
6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of containment as required by 10 CFR Part 50.54(o) and 10 CFR Part 50. Appendix J, Option B., as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995 and ANSI/ANS-56.8-1994.
The peak calculated containment internal pressure for the design basis loss-of coolant accident, P,, is 52 psig.
The maximum allowable containment leakage rate, L,, at P. shall be 0.1X of containment air weight per day.
Leakage rate acceptance criteria are:
a.
Containment leakage rate acceptance criterion is ~ 1.0 L.
During the first unit startup following testing in accordance with this.
rogram. the leakage rate acceptance criteria are ~ 0.6 L, for the ype B and Type C tests and s 0.75 L, for Type A tests; b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is ~ 0.05 L, when tested at a P,,
PALO VERDE - UNIT 3 6-25 Amendment No. 84-M. 85
DNINISTRATIV CONTRO~II
~aeaawaatLIelnaaaaaaaa
(
aa d) 2)
For each door, leakage rate is;g 0.01 L, ashen pressurized to
.h,14.5 k Oe5 ipsig.
L PAI.O VERDE - UNIT 3 6-26 Amendm'e.nt Nn. 75
'