Text

Proposed Tech Specs,Allowing Implementation of Recently Approved Option B to 10CFR50,App J
	ML20101R405
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/11/1996
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20101R404	List: 0CAN049602, Application for Amends to Licenses DPR-51 & NPF-6,allowing Implementation of Recently Approved Option B to 10CFR50,App J; ML20101R405;
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	NUDOCS 9604160262
	Download: ML20101R405 (44)
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l ATTACHMENT ID OCAN049602 PROPOSED TECHNICAL SPECIFICATION AND RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING l

LICENSE NOs. DPR-51 and NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNITS ONE & TWO DOCKET NOs. 50-313 and 50-368 l

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9604160262 960411 PDR ADOCK 05000313 1

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O Attachment to OCAN049602 Page 1 of 6 DESCRIPTION OF PROPOSED CHANGES

1. Description of proposed ANO-1 changes:

Changed Technical Specification (TS) 4.4.1.1, 4.4.1.1.4, 4.4.1.2, and 4.4.1.2.5 to require leakage rate tests to be conducted in accordance with the Reactor Building Leakage Rate Testing Program.

Relocated the applicable information from TS sections 4.4.1.1.1, 4.4.1.1.2, 4.4.1.1.3, 4.4.1.1.5, 4.4.1.1.6, 4.4.1.1.7, 4.4.1.2.1, 4.4.1.2.2, 4.4.1.2.3, 4.4.1.2.4, 4.4.1.2.5, 4.4.1.3, and 4.4.1.5 to the Reactor Building Leakage Rate Testing Program. The information in these Specifications that was not allowed under Option B was removed.

Revised bases information in TS 4.4.1 to be consistent with the Reactor Building Leakage Rate Testing Program.

Placed all of Specification 4.4.1 on page 79, and placed its bases on page 80, for human factors considerations.

Added section 6.8.4 which requires the Reactor Building Leakage Rate Testing Program.

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Description of proposed ANO-2 changes:

Added section 6.15, Containment Leakage Rate Testing Program, to the index page.

Modified Specification 4.6.1.1.c to require leak rate testing of the equipment hatch seal in accordance with the Containment Leakage Rate Testing Program. A typographical error was also corrected in the title for section 3/4.6.1.

Modified Specification 3/4.6.1.2 to insure the containment leakage rates are in accordance with the Containment Leakage Rate Testing Program. Also removed the limits that were repetitive to the program.

The smveillance requirement 4.6.1.3.1 was modified to eliminate information from the Specification that exists in the Containment Leakage Rate Testing Program and a reference to the program was added.

The footnote concerning the surveillance requirement for air lock interlock was modified in accordance with the Improved Standard Technical Specifications (ISTS). These changes resulted in the renumbering of the remaining footnotes.

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Attachment to OCAN049602

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Page 2 of 6 The inspection requirements from TS 4.6.1.5.3 were removed because Option B requires e-the same inspections to be performed. The bases for this specification was changed to reflect the most current maximum containment pressure in the event of a loss of coolant accident.

Modified TS 3/4.6.1.2 bases to explain the leakage acceptance criteria and eliminated the e

information regarding low pressure testing of the containment due to no longer being

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allowed by Option B.- In addition, a reference to Option B of 10 CFR 50 Appendix J was added for clarity.

a The bases information for 3/4.6.1.3 was expanded by adding applicable bases information e

from the ISTS that could be used to help clarify the Specification and removed the old j

bases information that would be repetitive.

Added 6.15 to the Administrative Controls section requiring the Containment Leakage Rate Testing Program.

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1 BACKGROUND i-The Nuclear Regulatory Commission has amended its regulations to provide a performance based alternative for leakage rate testing of containments. The new testing alternative is 3

Option D of 10 CFR 50, Appendix J, and is available in lieu of compliance with the present prescriptive requirements contained in Option A of Appendix J.Section V.B. of Option B requires licensees who wish to voluntarily adopt Option B, or parts thereof, to submit to the l

NRC an implementation plan and a request for a revision to the Technical Specifications.

i Therefore, ANO is proposing appropriate TS changes to adopt Option B of 10 CFR 50 Appendix J with our implementation plan listed below, i

IMPLEMENTATION PLAN The Containment Leakage Rate Testing Program, as required by Option B of 10 CFR 50 Appendix J, and as identified by Section 6.8.4 and 6.15 of the proposed ANO-1 and ANO-2 1

Technical Specifications, will be effective prior to implementation of these amendments. The performance based leakage rate testing program will be developed consistent with Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.

DISCUSSION OF CHANGE ANO has committed to implementation of the ISTS, NUREG-1430, " Standard Technical i

Specifications Babcock and Wilcox Plants," and NUREG-1432, " Standard Technical Specifications Combustion Engineering Plants." The industry is currently working with the NRC to include Option B in the ISTS. The proposed changes for both units are believed to

~ be in accordance with the latest draft of Option B for the ISTS. In accordance with the criteria of 10 CFR 50.36, portions of the prescriptive information concerning leakage rate i

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i Attachment to OCAN049602

. Page 3 of 6 l-testing have been relocated out of the Technical Specifications and a leakage rate testing program is established and referenced in the Administrative Controls Section. The proposed changes to ANO-2 TS 3/4.6.1.3 are based on the approval of the proposed amendment described in letter 2CAN049511.

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1. Discussion of proposed ANO-1 changes:

i Specifications 4.4.1.1,4.4.1.1.4,4.4.1.2, and 4.4.1.2.5 have been changed to require Leakage Rate Tests to be conducted in accordance with the Reactor Building Leakage Rate Testing Program. This change is in accordance with the ISTS and is considered administrative.

j The applicable information from TS sections 4.4.1.1.1, 4.4.1.1.2, 4.4.1.1.3, 4.4.1.1.5, 4.4.1,1.6, 4.4.1.1.7, 4.4.1.2.1, 4.4.1.2.2, 4.4.1.2.3, 4.4.1.2.5, 4.4.1.3, and 4.4.1.5 has been relocated to the Reactor Building Leakage Rate Testing Program in accordance with ISTS.

The information in these Specifications that was not allowed under Option B was removed.

l The design pressure of the Reactor Building (59 psig) was previously assumed to be the same value used for P., Under Option B testing, P. has been clearly defined as the peak calculated 1

internal pressure related to the design basis loss of coolant accident. The most recent revision to this calculation reflects that 53.96 psig is the design bases loss of coolant accident reactor building peak pressure. The value of 54 psig was chosen for conservatism over the value of 53.96 psig for this change. Therefore, P. was corrected to 54 psig and is listed in Section 4

6.8.4 of the proposed change.

l The bases information under TS 4.4.1 has been modified to be consistent with the Reactor 4

Building Leakage Rate Testing Program. The information regarding testing prior to initial operation was also removed because the information is located in the SAR. Also added bases information for this Specification that explains the reactor building leakage rate acceptance criteria in accordance with the ISTS.

Section 6.8.4 was added to require the Reactor Building Leakage Rate Testing Program. This section is in accordance with the latest draft of the ISTS with the exception of the air lock acceptance criteria. The air locks are tested penetrations that require Type B tests. Our current TS include any leakage from the air locks to be included in Specification 4.4.1.2.3.

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This is because the air locks are tested penetrations. The acceptance criteria located in TS 4.4.1.2.3 states "the total leakage from all tested penetrations and isolation valves shall not exceed 60% L.". Section 6.8.4 of the proposed change maintains the requirement for the air locks to be Type B tested with the same acceptance criteria of s 0.60 L. for the total leakage from all Type B and Type C tests.

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Attachment to

.a OCAN049602 Page 4 of 6 2.

Discussion of proposed ANO-2 changes:

The surveillance requirement 4.6.1.3.1 was modified to eliminate information that exists in the Containment Leakage Rate Testing Program and a reference to the program was added. The j

associated footnotes were renumbered and the footnote concerning the surveillance requirement for air lock interlock was modified in accordance with the ISTS.

The prescriptive requirements from 4.6.1.1.c and 3/4.6.1.2 were relocated to Section 6.15 and 4

the low pressure testing requirements that are not allowed by Option B have been removed.

The prescriptive surveillance frequencies were also removed from these specifications in order to adopt the new performance based testing frequencies allowed by Option B. These changes

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are to make the TS consistent with drafted Option B testing requirements to be implemented in the ISTS.

The inspection requirements from 4.6.1.5.3 were removed due to the these requirements being included in Option B and therefore included in the Containment Leakage Rate Testing q

Program. The relocation of this requirement out of the TS is consistent with the ISTS. The bases for this specification was also changed to reflect the most current maximum containment pressure of 54 psig in the event of a loss of coolant accident. This pressure is acceptable due j

to the design pressure of the containment building being 54 psig.

The bases information for 3/4.6.1.2 was modified to explain the leakage rate acceptance criteria and to eliminate the information regarding low pressure testing of the containment due to no longer being allowed by Option B. The added bases for the leakage rate acceptance criteria is in accordance with the ISTS. A reference to Option B of 10 CFR 50 Appendix J was also added.

The bases information for 3/4.6.1.3 was expanded by adding bases information from the ISTS that could be used for clarity and removed the information that would be repetitive. These changes are in accordance with the ISTS.

Section 6.15 was added to the Administrative Controls Section requiring the Containment l

Leakage Rate Testing Program. This program is in accordance with the latest ISTS draft r

i Other administrative changes that are being proposed with this change is addition of Section 6.15, Containment Leakage Rate Testing Program on page XVII of the TS index page and the correction of the spelling of containment in the heading of 3/4.6.1.

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Attachment to OCAN049602 Page 5 of 6 DETERMINATION OF NO SIGNIFICANT IIAZARDS CONSIDERATION An evaluat. ion of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92.(c). A discussion of these standards as they relate to this amendment request follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed changes to the Technical Specifications implement Option B of 10 CFR 50 Appendix J at ANO. The proposed changes will result in increased intervals between containment leakage tests determined through a performance based approach. The intervals between such tests are not related to conditions which cause accidents. The proposed changes do not involve a change to the plant design or operation. Therefore, this change does n01 involve a significant increase in the probability of any accident previously evaluated.

NUREG-1493, " Performance-Based Containment Leak-Test Program," contributed to the technical bases for Option B of 10 CFR 50 Appendix J. NUREG-1493 contains a detailed evaluation of the expected leakage from containment and the associated consequences. The increased risk due to lengthening of the intervals between containment leakage tests was also evaluated and found acceptable. Using a statistical approach, NUREG-1493 determined the increase in the expected dose to the public from extending the testing frequency is extremely small. It also concluded that a small increase is justifiable due to the benefits which accrue from the interval extension. The primary benefit is in the reduction in occupational exposure.

The reduction in the occupational exposure is a real reduction, while the small increase to the public is statistically derived using conservative assumptions. Therefore, this change does Dgt involve a significant increase in the consequences of any accident previously evaluated.

Therefore, this change does Uni involve a significam increase in the probability or consequences of any accident previously evaluated.

Attachment to OCAN049602 Page 6 of 6 i

Criterion 2-Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

The proposed change to the Technical Specifications incorporates the performance based approach authorized by Option B of 10 CFR 50 Appendix J. The interval extensions allowed l

by this change do not involve a change to the plant design or operation. No safety related j

equipment or safety functions are altered as a result of this change. The reduced testing frequency does not affect the testing methodology. As a result, the proposed change does not affect any of the parameters or conditions that could contribute to initiation of any accidents.

No new accident modes are created by extending the test intervals. Therefore, this change does nol create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 - Does Not lavolve a Significant Reduction in the Margin of Safety.

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The proposed change does not change the performance methodology of the containment leakage rate testing program. However, the proposed change does affect the frequency of containment leakage rate testing. With an increased frequency between tests, the proposed j

change does increase the probability that a increase in leakage could go undetected for a longer period of time. Operational experience has demonstrated the leak tightness of the containment buildings has been significantly below the allowable leakage limit.

4 The margin to safety that has the potential of being impacted by the proposed change involves i

the offsite dose consequences of postulated accidents which are directly related to containment leakage rates. The limitation on containment leakage rate is designed to ensure the total leakage volume will not exceed the value assumed in our accident analysis. The j

margin to safety for the offsite dose consequences of postulated accidents directly related to containment leakage is maintained by meeting the 1.0 L. acceptance criteria. The proposed j

change maintains the 1.0 L acceptance criteria.

Therefore, this change does nat involve a significant reduction in the margin of safety.

4 Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does nol 4

involve a significant hazards consideration.

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PROPOSED TECHNICAL SPECIFICATION CHANGES i

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ANO-1 i

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REACTOR BUILDING 4.4.1 Reactor Building Leakage Tests

. Applicability 4

Applies to the reactor building.

objective 1

i To verify that leakage from the reactor building is maintained within allowable limits.

Specification a

4.4.1.1 Integrated leakage rate tests shall be conducted in accordance with the Reactor Building Leakage Rate Testing Program.

4.4.1.1.1 Deleted l

4.4.1.1.2 Deleted l

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4.4.1.1.4 Integrated leakage rate testing frequencies shall be in accordance f

with 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

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4.4.1.2.5 Local leakage rate testing frequencies shall be in accordance with the Reactor Building Leakage Rate Testing Program.

1 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

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' Bases (1)

The reactor building 's designed for an internal pressure of 59 psig and a steam-air mixture temperature of 285'F.

I The peak calculated reactor building pressure for the design basis loss of coolant accident, Pa, is 54 psig. The maximum allowable reactor building

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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.

1 REFERENCE (1) FSAR, Sections 5 and 13.

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6.8.2 Each procedure of 6.8.1 abova, and changas in intent thersto, 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:

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The intent of the original procedure is not altered.

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The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license on Unit 1.

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The change is documented, reviewed and approved as required -

by the QAMO, within 14 days of implementation.

6.8.4 The Reactor Building Leakage Rate Testing Program shall be established, implemented, and maintained:

A program shall be established to implement the leakage rate testing i

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 maximum 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 l

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.

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION................................

6-12 6.7 SAFETY LIMIT VIOLATION.................................

6-13 6.8 PROCEDURES.............................................

6-13 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 ENVIRONMENTAL OUALIFICATION..........................

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.............

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d ARKANSAS - UNIT 2 XVII Amendment No. M,60,M,M,4M,

1 3/4.6 CONTAINMENT SYSTEM

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3/4.6.1 PRIMARY CONTAINMENT l

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CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT S5ANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 permitted by Specification 3.6.3.1.

b.

By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

c.

After each closing of the equipment hatch, by leak rate testing 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,

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 lindts, prior to increasing the Reactor Coolant System temperature above 200*F.

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 6-3 Amendment No, i

i CONTAINMENT SYSTEMS-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 OPERABLE by testing the air lock interlock mechanism at least once per 184 i

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'An inoperable air lock door does not invalidate the previous successful I

performance-of the overall air lock leakage test, j

'This surveillance requirement is only required to be performed upon entry i

or exit through the associated containment air lock.

l ARKANSAS - UNIT 2 3/4 6-5 Amendment No.

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CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION 4.6.1.5.2 End Anchorages and Adjacent Concrete Surfaces The structural integrity of the.end anchorages of all tendons" inspected pursuant to Specification 4.6.1.5.1 and the adjacent concrete surfaces shall be demonstrated by determining 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.

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ARKANSAS'- UNIT 2 3/4 6-9 Amendment No. 41,444,

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3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY 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 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 accident analyses at the peak design basis loss of coolant accident pressure, Pa, of 54 psig. As an added conservatism, the measured overall integrated leakage rate is further limited to s 0.75 La during the performance of the periodic tests to account for possible degradation of the containment leakage barrisrs 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 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.

3/4.6.1.3 CONTAINMENT AIR LOCKS Each containment 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 mitigation 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 e 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.

ARKANSAS - UNIT 2 B 3/4 6-1 Amendment No.

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CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE, AIR TEMPERATURE AND RELATIVE HUMIDITY The limitations on containment internal pressure, average air temperature and relative humidity 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 tempe ature ensures that the containment liner plate temperature does not exceed the d. sign temperature of 300*F during LOCA conditions. The containment temperature lindt 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 li.te of the facility.

Structural integrity is required to ensure that the containment will withstand the maximum pressure of 54 psig in j

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 1 containment are sufficient to d4monstrate 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 *xhaust 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. 449,

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 air weight per day at Pa.

Leakage rate acceptance criteria ares a.

Containment 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.

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 6-26 Amendment No.

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MARKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY)

4.4 REACTOR BUILDING

' d 4.4.1 Eeactor Buildino Leakace Tests Acolicability 4

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Specification 4

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-..u..g 4.4.1.1.4 Tr c.;r. v c' T::tInteorated leakaae rate l

Tiesting frequencies shall be in accordance with 100""50, L..

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Buildino Leakaae Rate Testino Procram.

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i Amendment No. 141, 82

4.4.1.2.5 T :t F::q::ncy Local leakace rate testina freauencies shall be in accordance with the Reactor Buildina Leakaa,e Rate Testina Procram.

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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 y

practical during plant operation. The latter valves shall be s

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2 1

l Amendment No. M,M, 83

l th: d;;ig. p::: urs.

Th::: t :t: uill v: kfy that th:-leck:g: ::t f:::

e:::ter building p::::uri: tion ::tisfic th: : lation: hip given in the epeeff4eet-lene The perfeemenee-ef : pariedic int grated lenhage ::t test during plant life p;cvid::

cur nt :::::: ent :f pet-entiel leck:g fc th: ::::ter i

building in :::: Of sn-eeeldent-that xculd-pres:uri:: the interier ;f-the

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building icel: tion v 14ee-er; t: be cle::d in th; norm:1 nner.

TM -teet p::::ur: Of 30 p:ig for the periodic integrated le: hag rate test 1:

eufficiently high te provid; an-eeeurst: meccurement of the 10: hag: :st:

and it duplic te: the p : p ::tional-keakag: :st: t::t et 30 p:ig.

The

p::ification pr;vid:: : relati:n: hip for : leting the mes:ured le hag: ef ei: at 30 p;ig to the p;tanti:1 leakage-et 50 prig.

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First is the leu probability of leck in the liner, b : cue: ef confermance of th: :: plet ::::ter building t:

0.201 10 hag: ::t t 50 p:ig during p : Operational t sting and th:

abeeeee-ef any significent stresses in th; liner during ::: ter speration S : nd is th: :::: frequent-t:: ting, et d::ign p :: ure, of th::: pertion:

ef-ehe-eee ter bui4 ding :nvel;p that ::: :::t likely to develop lecke during ::::ter sp : tien (penet-ra ti on :nd i::lati n valve:)

nd the low v:1ue ef.50L 10:hage that i: sp;cified : ::::pt:ble f:;r te:ted p;nctration: :nd i:01stien valves.

Third is the tenden streee-eueveillen;;

progrcr which p :vid:: ::euvence that en imp : tant p:et of th: structural inte;:ity :f the reactee-bui4 din; i: : 1nt-einedr The peak calculated reactor buildino pressure for the desian basis loss of coolant accident, Pac-is 54 osia. The maximum allowable reactor buildina leakaae rate, La, sHall be 0.20% of containment air weicht per day at Pas The reactor buildine will be periodically leakace tested in accordance with the Reactor Buildina Leakace Rate Testina Procram.

These periodic testina recuirements verify the reactor buildina leakace rate does not exceed the assumptions used in the safety analysis. At 5 1.0 La the offsite dose consecuences are bounded by the assumptions of the s'afety analysis.

Durina the first unit startup followina testina in accordance with this procram, the leakace rate acceptance criteria are s 0.60 La for the combined Type B and Type C leakage, and 5 0.75 La for overall Tvoe A leakace. At all other times between recuired leaka_ce t'ests. the acceptance criteria is based on an overall Type A leakace limit of s 1.0 La-REFERENCE l

(1) FSAR, Sections 5 and 13.

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4 84

1 6.8;2 Each procsdure of 6.8.1 above, and changas in intent thareto, 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 management staff, at least one of whom holds a Senior Reactor Operator's license on Unit 1.

c.

The change is documented, reviewed and approved as required by the QAMO, within 14 days of implementation.

6.8.4 The Reactor Buildino Leakaae Rate Testino Procram shall be established, implemented, and maintained:

A procram shall be established to implement the leakaae rate testing of the reactor buildino as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This procram shall be in accordance with the cuidelines contained in Reculatory Guide 1.163, " Performance-Based Containment Leak-Test 4

Procram," dated September 1995.

The peak calculated reactor buildina internal pressure for the l

desian basis loss of coolant accident, Pa, is 54 osio.

The maximum allowable reactor buildina leakace rate, La, shall

~

be 0.20% of containment air weicht per day at P g3 Reactor buildina leakace rate acceptance criteria is s 1.0 Laa Durina the first unit startup followina each test performed In accordance with this procram, 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 frecuencies specified in the Reactor Buildina Leakace Rate Testino Prooram.

The provisions of Specification 4.0.3 are applicable to the Reactor 4

Buildina Leakace Rate Testino Procram.

t 1

Amendment No. 46,40,34,47,74,84, 127 i

BB,94,99,409,444,444, 44-7, M9, M&, M9, i

O e

MARKUP OF CURRENT ANO-2 TECHNICAL SPECIFICATIONS (FOR INFO ONLY)

4 E

INDEX o,

)*

ADMINISTRATIVE CONTROLS i

SECTION PAGE 6.6 REPORTABLE EVENT ACTION................................

6-12 j

6.7 SAFETY LIMIT VIOLATION.................................