ML20128H479
| ML20128H479 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/03/1996 |
| From: | Alexion T Office of Nuclear Reactor Regulation |
| To: | Entergy Operations |
| Shared Package | |
| ML20128H484 | List: |
| References | |
| DPR-51-A-185, NPF-06-A-176 NUDOCS 9610090361 | |
| Download: ML20128H479 (25) | |
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@ REati ch UNITED STATES s
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2064 5 0001
\\...../
ENTERGY OPERATIONS. INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 185 License No. DPR-51 1.
Tte Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated April 11, 1996, as supplemented August 23, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and 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 withcut 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 license 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.
9610090361 961003 PDR ADOCK 05000313 p
. 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-6 is hereby amended to read as follows:
(P) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.185, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION G
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'C M Thomas W. Alexion, Project ' Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 3, 1996 1
ATTACHMENT TO LICENSE AMENDMENT NO. 185 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 I
Revise the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES 79 79 80 80 81 81 82 82 83 83 84 84 127 127 4
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4.4 REACTOR BUZLDING 4.4.1 Reactor Building Leakage Tests l
Applicability Applies to the reactor building.
Objective 1
To verify that leakage from the reactor building is maintained within allowable limits.
Specification 4.4.1.1 Integrated leakage rate tests shall be conducted and visual inspections performed in accordance with the Reactor Building i
Leakage Rate Testing Program.
4.4.1.1.1 Deleted l
4.4.1.1.2 Deleted l
4.4.1.1.3 Deleted l
f 4.4.1.1.4 Integrated leakaga rate testing frequencies shall be in accordance Nith the Reactor Building Leakage Rate Testing Program.
4.4.1.1.5 Deleted l
4.4.1.1.6 Deleted l
4.4.1.1.7 Deleted l
4.4.1.2 Local leakage rate tests shall be conducted in accordance with the Reactor Building Leakage Rate Testing Program.
4.4.1.2.1 Deleted l
4.4.1.2.2 Deleted l
4.4.1.2.3 Deleted l
4.4.1.2.4 Deleted l
4.4.1.2.5 Local leakage rate testing frequencies shall be in accordance with the Reactor Building Leakage Rate Testing Program.
4.4.1.3 Deleted l
4.4.1.4 Isolation Valve Functional Tests Every three months, remotely operated reactor building isolation valves shall be stroked to the position required to fulfill their safety function unless such operation is not practical during plant operation. The latter valves shall be tested once every 18 months.
4.4.1.5 Deleted l
Amendment No. H,n,m,m,m,185 79
Bases (1)
The reactor building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 285'F.
The peak calculated reactor building pressure for the design basis loss of coolant accident, Pa, is 54 psig. The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.
The reactor building will be periodically leakage tested in accordance with the Reactor Building Leakage Rate Testing Program. These periodic testing requirements verify the reactor building leakage rate does not exceed the assumptions used in the safety analysis. At s 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C leakage, and s 0.75 La for overall Type A leakage. At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 La.
REFERENCE (1) FSAR, Sections 5 and 13.
i Amendment No. m, m,4M,185 80
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Amendment No.185 el
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Amendment No an,185 82
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Amendment No. u,M,185 83
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Amendment No. 185 84
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6.8.2 Each procadure of 6.8.1 tbova, and changas in intent thorato, shall be reviewed and approved as required by the QAMO prior to implementation-and reviewed periodically as set forth in administrative procedures.
6.8.3 Changes to procedures of 6.8.1 above may be made and implemented prior to obtaining the review and approval required in 6.8.2 above provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two members of the plant nanagement staff, at least one of whom holds a Senior Reactor Operator's license on Unit 1.
The change is documented, reviewed and approved as required c.
by the QAMO, within 14 days of implementation.
6.8.4 The Reactor Building Leakage Rate Testing Program shall be established, inplemented, and maintained:
A program shall be established to implement the leakage rate testing of the reactor building as required by 10 CFR 50.54 (o) and 10 CFR 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.
The peak calculated reactor building internal pressure for the design basis loss of coolant accident, Pa, is 54 psig.
The umximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.
Reactor building leakage rate acceptance criteria is s 1.0 La.
During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the Type B and Type C tests and s 0.75 La for Type A tests.
The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.
The provisions of Specification 4.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.
Amendment No. M,M, M,M,M,M, 127 M, M, M, M9, M4, M3, se,u9,M6,4M,185
9e nau Jt UNITED STATES y
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 206MM001 gs...../
INTERGY OPERATIONS. INC.
DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.176 License No. NPF-6 1.
The Nuclear Regulatory Comissiot6 (the Comission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated April 11, 1996, as supplemented August 23, 1996, complies wfth the standards and requirements of the Atomic Energy Act of 1951,, as amended (the Act), and the Comission's rules and regulatiors 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 Comission; C.
There is reasonable assurance:
(1) 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 license 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.
. 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. DPR-51 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.176, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jw,#
b,',
Thomas W. Alexion, Project anager a
Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 3, 1996
ATTACHMENT TO LICENSE AMENDMENT NO.176 FACILITY OPERATING LICENSE N0. NPF-6 DOCKET NO. 50-368 Revise the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMQVE PAGES INSERT PAGES XVII
. XVII 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-5 3/4 6-5 3/4 6-9 3/4 6-9 B 3/4 6-1 B 3/4 6-1 B 3/4 6-la B 3/4 6-2 B 3/4 6-2 6-26 4
i
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION................................
6-12 6.7 SAFETY LIMIT VIO1ATION.................................
6-13 6.8 PROCEDURES.............................................
6-13 l
l 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS..................................
6-14 6.9.2 SPECIAL REPORTS..................................
6-16 6.9.3 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT...
6-18 6.9.4 ANNUAL RADIOLOGICAL ENVIRONMENT OPERATING REPORT.
6-20 6.9.5 CORE OPERATING LIMITS REPORT.....................
6-21 6.10 RECORD RETENTION.....................................
6-22 6.11 RADIATION PROTECTION PROGRAM.........................
6-23 6.12 ENVI RONMENTAL OUALI FI CATION..........................
6-23 6.13 HIGH RADIATION AREA.................................
6-24 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)...............
6-25 6.15 CONTAINMENT LEAKAGE RATE TESTING PROGRAM.............
6-26 l
i ARKANSAS - UNIT 2 XVII Amendment No M,60,M,94,W,176
3/4. 6 CONTAINKENT SYSTEM 3/4.6.1 PRIMUtY CONTAINMENT l
CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
j APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour 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 REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
a.
At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except for valves that are open under administrative control as perndtted by Specification 3.6.3.1.
b.
By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.
After each closing of the equipment hatch, by leak rate testing c.
the equipment hatch seals in accordance with the containment Leakage Rate Testing Program.
- Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need_ not be performed more often than once per 92 days.
ARKANSAS - UNIT 2 3/4 6-1 Amendment No. M4,176 4
CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 containment leakage rates shall be in accordance with the containment Leakage Rate Testing Program.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With containment leakage rates not within limits, restore containment leakage to within limits, prior to increasing the Reactor Coolant System temperature above 200*r.
SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be determined in accordance with the Containment Leakage Rate Testing Program.
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ARKANSAS - UNIT 2 3/4 6-2 Amendment No.176
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ARKANSAS - UNIT 2 3/4 6-3 Amendment No.176
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE.
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APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
I With one containment air lock door inoperable in one or more containment a.
air locks*:
1.
Verify that at least the OPERABLE air lock door is closed in the affected air lock within one hour 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'.
2.
Operation may then continue provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.
3.
Otherwise, be in at least HOT STANDBY within the next six 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 />, b.
With the containment air lock interlock inoperable in one or more 2
containment air locks :
1.
Verify that at least one OPERABLE air lock door is closed in the af fected air lock within one hour and restore the inoperable air lock interlock 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 an d
OPERABLE air lock door closed.
2.
Operation may then continue provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.
3.
Otherwise, be in at least HOT STANDBY within the next six 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 />.
With one or more air locks inoperable for reasons other than those c.
addressed in ACTION a. or b.:
1.
Immediately initiate action to evaluate overall containment leakage per LCO 3.6.1.2.
2.
Verify that at least one door in the affected air lock is closed within one hour and restore the affected 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 />.
3.
Otherwise, be in at least HOT STANDBY within tbe next six 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 />.
' Separate ACTION entry is allowed for each air lock.
With both air locks inoperable, entry and exit is permissible for seven days 2
under administrative controls.
Entry and exit is pernissible to perform repairs on the af fected air lock 8
components.
' Entry and exit is permissible under the control of a dedicated individual.
ARKANSAS - UNIT 2 3/4 6-4 Amendment No. 175
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CONTAZNMENT SYSTEMS j
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SURVEILLANCE REQUIREMENTS l
4.6.1.3.1 Each containment air lock shall be demonstrated OPERABLE as specified in the Containment Leakage Rate Testing Program.
4.6.1.3.2 Each containment air lock interlock shall be demonstrated CPERABLE by testing the air lock inter 3ock mechanism at least once per 184 days'.
1
- Leakrate results shall also be evaluated against the acceptance criteria of specification 3.6.1.2.
' An inoperable air lock door does not invalidate ?.he previous successful performance of the overall air lock leakage test.
This surveillance requirement is only required to be performed upon entry or exit through the associated containment air lock.
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l ARKANSAS - UNIT 2 3/4 6-5 Amendment No. i?5-176
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CONTAINMENT SYSTEMS INTERNAL PRESSURE, AIR TEMPERATURE AND RELATIVE HUMIDITY LIMITING CONDITION FOR OPERATION 3.6.1.4 The combination of containment internal pressure, average air temperature and relative humidity shall be maintained within the region of acceptable operation shown on Figure 3.6-1.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the point defined by the combination of containment internal pressure, average air temperature and relative humidity outside the region of acceptable operation shown on Figure 3.6-1, restore the combination of containment internal pressure, average air temperature and relative humidity to within the above limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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 REQUIREMENTS 4.6.1.4 The primary containment internal pressure, average air temperature and relative humidity shall be determined to within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The containment average air temperature shall be the temperature of the air in the containment HVAC common return air duct upstream of the fan / cooler units.
ARKANSAS-UNIT 2 3/46-6
CONTAIRMENT SYSTEMS LIMITING CONDITION FOR OPERATION 4.6.1.5.2 End Anchorages and Ad$acent Concrete surfaces The structural integrity of the end anchorages of all tend 7ns inspected pursuant to specification 4.6.1.5.1 and the adjacent concrete surfaces shall be demonstrated by detennining through inspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjacent to the end anchorages. Inspections of the concrete shall be performed during the Type A containment leakage rate tests (reference specification 4.6.1.2) while the containment is at its maximum test pressure.
4.6.1.5.3 containment surfaces The structural integrity of the exposed acces sible interior and exterior surfaces of the containment, including the liner plate, shall be determined by a visual inspection of these surfaces and verifying no apparent changes in appeata1ce or other abnormal degradation has occurred in accordance with the Containment Leakage Rate Testing Program.
l ARKANSAS - UNIT 2 3/4 6-9 Amendment No. M,MS,176
l 3/4.6 CONTAINMENT SYSTEMS l
BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY,
Primary CONTAINMENT INTEGRITY ensures that the release of radioactive
)
l materials from the containennt atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the j
accident analyses at the peak design basis loss of coolant accident pressure, Par j
of 54 psig. As an added conservatism, the measured overall integrated leakage rate is further limited to 5 0.75 La during the 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 50.
l The containment will be periodically leakage tested in accordance with the Containment Leakage Rate Testing Program. These periodic testing requirements verify the containment leakage rate does not exceed the assumptions used in the safety analysis. At s 1.0 La rhe offsite dose consequences are bounded by the assumptions of the safety analysis, i
I During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C leakage, and s 0.75 La for overall Type A leakage. At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 La.
3/4.6.1.3 CONTAINMENT AIR LOCKS Each contain; nt air lock forms part of the containment pressure boundary.
As part of the containment, the air lock safety function is related to control of the containment leakage rate resulting from a DBA.
Thus, each air lock's structural integrity and leak tightness are essential to the successful ndtigation of such an event.
For the purposes of this specification, the vertical end plates of the air lock barrel, on which the doors themselves are mounted, shall be considered part of the door.
Each air lock is required to be OPERABLE.
For the air lock to be considered OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.
Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.
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ARKANSAS - UNIT 2 B 3/4 6-1 Amendment No. 146,176
3/4.6 CONTAINMENT SYSTEMS BASES Action statement "a" has been modified by a note.
Note 2 allows use of the air lock for entry and exit for 7 days under administrative controls if both air locks have an inoperable door. This 7 day restriction begins when the second air lock is discovered inoperable. Containment entry may be required to perform Technical Specification (TS) surveillances and actions, as well as other activities on equipment inside containment that are required by TS or activities on equipment that support TS required equipment. An example of such an activity would be the isolation of a containment penetration by at least one OPERABLE valve, and the subsequent repair and post maintenance TS surveillance testing en the inoperable valve.
In addition, containment entry may be required to perform repairs on vital plant equipment, which if not repaired, could lead to a plant transient or reactor trip. This note is not intended to preclude performing other activities (i.e.,
non-TS required activities or repairs on non-vital plant equipment) if the containment is entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probability of an event that could pressurire the containment during the short time that the OPERABLE door is expected to be open.
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l ARKANSAS - UNIT 2 B 3/4 6-la Amendment No. 4-%,176
a CONTAINMENT SYSTEMS RASES 3/4.6.1.4 INTERNAL PRESSURE, AIR TEMPERATURE AND RELATIVE HUMIDITY The limitations on containment internal pressure, average air temperature and relative hmsidity ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 5.0 psig, 2) the containment peak pressure does not exceed the design pressure of 54 psig during design basis conditions, and 3) the ECCS analysis assumptions are maintained.
The limitation on containment average air temperature ensures that the containment liner plate temperature does not exceed the design temperature of 300'T during LOCA condi* ions. The containment temperature limit is consistent with the accident analyses.
Figure 3.6-1 represents analysis limits and does not account for instrument error.
3/4.6.1.5 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of $4 psig in l
the event of a LOCA.
The visual examination of tendons, anchorages and containment surfaces and the Type A leakage tests of the Unit 2 containment in conjunction with the required surveillance activities of the Unit I containment are sufficient to demonstrate this capability.
The surveillance requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures", January 1976.
3/4.6.1.6 CONTAINMENT VENTILATION SYSTEM The containment purge supply and exhaust isolation valves are required to be closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA.
Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system.
ARKANSAS - UNIT 2 B 3/4 6-2 Amendment No. H9,176
ADMINISTRATIVE CONTROLS 6.15 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 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.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 54 psig.
The maximum allowable containment leakage rate, La, shall be 0.1% of containment i
air weight per day at Pa.
i Leakage rate acceptance criteria are Containment leakage rate acceptance criteria is s 1.0 La.
During the a.
first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the Type B and Type C tests and 5 0.75 La for Type A tests.
b.
Air lock acceptance criteria are:
1.
Overall air lock leakage rate is s 0.05 La when tested at 2 Pa.
2.
Leakage rate for each door is s 0.01 La when pressurized to 2 10 psig.
The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.
ARKANSAS - UNIT 2 6-T6 Amendment No.176