ML20141G529

Amend 184 to License NPF-6,changing Allowable primary-to-secondary Leak Rate & in SRs Section of TSs Changing Acceptance Criteria for SG Tubes

ML20141G529
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/20/1997
From:	Kalman G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20141G536	List: ML20141G529 ML20141G546
References
NPF-06-A-184 NUDOCS 9705220404
Download: ML20141G529 (11)

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UNITED STATES g?

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666-0001 -

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ENTERGY OPERATIONS. INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.184 License No. NPF-6 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Operations, Inc. (the licensee) dated November 26, 1996, as supplemented by letter dated February 12, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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; j

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 j

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.

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9705220404 970520 PDR ADOCK 05000368 P

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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:

2.

Technical Soecifications d

The Technical Specifications contained in Appendix A, as revised through Amendment No. 184,'are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the

. Technical Specifications.

3.

The license amendment is effective within 30 days of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Sy

+e George Kalman, Senior 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:

May 20, 1997 f

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i ATTACHMENT TO LICENSE AMENDMENT NO.184 FACILITY OPERATING LICENSE NO. 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 1

contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE PAGES INSERT PAGES 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-14 3/4 4-14 B 3/4 4-2 B 3/4 4-2 8 3/4 4-3 8 3/4 4-3 8 3/4 4-4 B 3/4 4-4 4

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REACTOR COOLANT SYSTEM SURVEILIANCE REQUIRDENTS (C:ntinutd) 4.4.5.4 Acceptance criteria a.

As used in this Specification 1.

Tubing or Tube means that portion of the tube or sleeve which forms the primary system to secondary system pressure boundary.

2.

Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as isperfections.

f 3.

Degradation means a service-induced cracking, wastage, wear or generai corrosion occurring on either inside or outside of a tube.

4.

Degraded Tube means a tube containing imperfections 2t20% of nominal wall thickness caused by degradation.

5.

4 Degradation means the percentage of.the tube wall thickness af fected or removed by degradation.

6.

Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube containing a defect is defective.

7.

Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the ne,.c inspection. The plugging or repair limit is equal to 40% of the nominal parent tube and sleeve wall thickness for sleeves installed in accordance with B&W Topical Report BAW-2045-PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Dasign -

Application to ANO Unit 2".

The plugging limit is equal to 29%

of the nominal sleeve wall thickness within the sleeve pressure boundary for sleeves installed in accordance with CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight Sleeves," Revision 01, dated November 1996.

8.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

9.

Tube Inspection sneans an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

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ARKANSAS - UNIT 2 3/4 4-9

. Amendment No. 484,444,184 i

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REACTOR C001 ANT SYSTEM SURVEILLANCE REQUIREMENTS (Continuid)

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Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by addy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER j

OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.

Defective tubes may be repaired in accordance with:

1)

BEW Topical Report RAN-2045PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design-Application to ANO Unit 2".

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2) CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight Sleeves," Revision 01, dated November 1996. The post weld heat treatment. described in CEN-630-P shall.be performed.

4.4.5.5 Reports Following each inservice inspection of steam generator tubes the I

a.

number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

b.

The complete results of the steam generator tube inservice inspection shall be reported on an annual basis for the period in which the inspection was completed. This report shall include:

1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

3.

Identification of tubes plugged or sleeved.

4 Results of steam generator tube inspections which fall into c.

Category C-3 shall be reported in a Special Report pursuant to specification 6.9.2 an denoted by Table 4.4-2.

Notification of i

the Commission will be made prior to resumption of plant e

operation. The written Special Report shall provide.a description of investigations conducted to determine cause of the tube I

degradation and corrective measures taken to prevent recurrence.

ARKANSAS - UNIT 2 3/4 4-10 Amendment No. M,W,W,184 e

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3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE l

LEAKAGE DETECTION SYSTEMS

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LIMITING CONDITION FOR OPERATION

3. a. 6.1 The following Reactor Coolant System leakage detection systems shall be OPERASt.E.

a.

A containment atmosphere particulata radioactivity monitoring

system, b.

The containment sump level monitoring system, and A containment atmos'phere gaseous radioactivity monitoring c.

system.

APPLICABILITY: MODES 1, 2. 3 and 4.

ACTION:

j With only two of the above required leakage detection systems OPERABLE.-

operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactivity monitor-4 ing system is inoperable; otherwise, 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

4. 4. 6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a.

Containment atmosphere particulate and gaseous monitoring systems-performance of CHANNEL CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3.

b.

Contairunent sump level monitoring system-perfomance of CHANNEL CALIBRATION at least once per 18 months.

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1ARKAMSAS - UNIT 2 3/4 4-13 a

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REACTOR COOLANT SYSTEM

- REACTOR COOLANT SYSTEM LEAKAGE e

LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor coolant System leakage shall be limited to:

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a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, 1

300 gs11ons per day total primary-to-secondary laakage through both c.

steam generators and 150 gallons per day through any one steam generator, d.

10 GPM IDENTIFIED LEAKAGE from the Aeactor Coolant System, and Leakage as specified in Table 3.4.6-1 for those Reactor Coolant

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System Pressure Isolation Valves identified in Table 3.4.6.1.

APPLICABILITY: MODES 1, 2, 3 and 4.

1 AC'I']N With any PRESSURI BOUNDARY LEAKAGE, be in at leest NOT STANDBY l

a.

within 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 />.

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the l

1eakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-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 SNUTDOWN within i

l the following 30 honrs.

With any Reactor Coolant System Pressure Isolation Valve leakage c.

greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two valves

• in each high pressure line having a non-functional valve and be in at least NOT 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 />.

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• These valves may include check valves for which the leakage rate has I

been verified, manual valves or automatic valves. Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.

ARKANSAS - UNIT 2 3/4 4-14 Amendment No. 184 ti: irt:i t!: !t

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s/4.4.1 nACTOR C00MRT UMys AND CDouurf CMIFMTICE -

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n. plant is designed to operste with both reactor ecolant amops.and associated reactor coolant pumps da operstian, and maintain IBtBI.above the f

admits specified by specifiestime 3.2.4 during all asemal e and j

anticipated transtests.

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"In 1KEE 3, a slagle reactor coolant loop provides sufficient best j

vemoval espebility for removing doesy beat; however, single. failure

-schaiderations greguire that two loops be OPERARE. ~'

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in 110 DES 4 and 3, ' single zesctor eoolant loop er sketdown cooling

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loop provides sufficient beat removal capability for removing decay heat; but single failure ecosiderettoms. require that at least two 3 oops be CPIKA33. Thus, if the reactor coolant loops are mot CPERABLE, -this..,

3;-*4ficatice requires suo abstdown ecoling loops to be CPI 2&RLE.,l,*,

The operation of ese Beactor Coolant Pump or ese shutdown cooling yesp provides adequate flow to ensure mixing, prevent stratificaties and produce gradual reactivity changes during boron concastratioc redoctions in the Reacter Coolant System. The reactivity change rate associated with d

boron reductions will, therefore, he within the capability of operator recognition and control.

3/4.4.2 and 3/4.4.31ATETY YALME *.

The pressurizer code safety valve.s aperate to prevent the.RCS,from being pressurised above its Safety 1.init of 2730 psia. dlach safety valve is designed to relieve 420,000 lbs. ger hour.of saturated steam at the valve setpoint. The relief capacity of a single 1 safety valve is adegaste t

to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERAB2, an operating abstdown cooling loop, connected to the RCS, provides overpressure relief

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-capability cod will prevent RCS everpressurisatico.

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During operation, all pressuriser code safety valves most be erERABLE to prevent the RCS from being pressurised above its safety limit of 2730 psia. The combined relief espacity -ef these valves is sufficient to limit the Reactor Coolant Systes pressure to within its Safety Limit of 2750 l,

psia following a complete loss of turbine generator IJed while operating j

at RATED TF.IRMAI POWER and assuming no reactor trip until the first Reactor Protective system trip setpoint (Pressuriser Pressure-Eigh) As

- reached (i.e., w credit.is taken for a direct reactor trip en the less of canrbine) and also assuming so operatiae of she stees deny walves.

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'13/44-1 doendment No. #,=e9-149 ARKAMSAS - UNIT 2

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REACTOR C00LANT SYSTEM BASES Demonstration of the safety valves' lift setting will occur only during shutdown and will be performed in accordance with the l

provisions of Section XI of the ASME Boiler and Pressure Vessel

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9 3/4.4.4 PRESsURIEER A steam bubble in the pressuriser ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressuriser code 3

safety valves against water relief. The steam bubble functions to relieve RCS pressure during all design transients.

The requirement that 150 KW of pressuriser heaters and their.

associated controls be capable of being supplied electrical power from an J

h emergency bus provides assurance that these heaters can be energized during a loss-of-offsite power condition to maintain natural circulation at NOT STAND 8Y.-

!i 3/4.4.5 STEAM GENERATORS

~ The Surveillance Requirements for -inspection of the steam generator T

tubes ensure that.the structural integrity of this portion of the RCS will be maintained. The program for in' service inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

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Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that i

there is evidence of mechanical damage or progressive degradation due to j

design, manufacturing errors, or inservice conditions that lead e.o i

corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken, j

The plant is expected to be operated in a manner such that the t

f secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the

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secondary coolant chemistry is not maintained within these limits, u~

. localized corrosion may likely result in-stress corrosion zracking. _

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The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary.iccolant l

system and the secondary coolant system (primary-to-secondary leakage

= 150 gallons per day per steam generator). Cracks having a primary-to-j

- secondary leakage less than this limit during operation will have an adequate 1

margin of safety to withstand the loads imposed during normal operation i

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and by postulated accidents. operating plants have demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors on the secondary system.

-Leakage in excess of this limit will require plant shutdown and an unscheduled inspection,-during which the leaking tubes will be located and plugged or repaired.

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k ARKANSAS - UNIT 2 3 3/4 4-2 Amendment No. 49,444;184 4

R% ACTOR COOLAIf? SYSTEM

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!j wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, I

' _ - _..._.it will be found during scheduled inservice steam generator tubes examinations..._

Plugging or sleeving will-be -required for-all-tubes with irgerfections l

exceeding the plugging or repair limit as defined in Surveillance Requirement l

4.4.5.4.a.

Defective tubes may be repaired by sleeving in accordance with the B&W Topical Report RAN-2045PA-00 as supplemented by the information

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provided in B&W Report 51-1212539-00, "8WNS Kinetic Sleeve Design-Application to ANO Unit 2" or CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator i

Tubes Using Leak Tight J1eeves," Revision 01, dated November 1996. Steam i

generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20b of the tube wall l

thickness. For sleeved tubes, the adequacy of the system that is used for periodic inservice inspection will he validated.. Additionally, mpgraded testing methods will be evaluated and appropriately implemented as better

j methods are developed and validated for commercial use.

I whenever the results of any steam generator tubing inservice inspection fall into Category C-3 certain results will be reported in a special Report to the Commission pursuant to specification 6.9.2 as denoted by Table 4.2-2.

Notification 4

of the cassaission will be made prior to. resumption of plant operation. Such cases-will be considered by the Cossaission on a case-by-case basis and may result in a o

zequirement for ai.alysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

1 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l*

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS i

The RCS leakage detection systems required by this specification are

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I provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the reconsnendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage 2

Detection Systems" May 1973.

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3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE

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is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is j

sufficiently low to ensure early detection of additional leakage.

)I The 10 GPM IDENTINIED LEAKAGE limitation provides allowances for a limited l

amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

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The surveillance Requirements for RCS Pressure Isolation Valves provide

. 4dded assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem 14CA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of

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_ the allowed limit.

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lI If ARKANSAS - UNIT 2 5 3/4 4-3 Amendment No. M,4M,444,184 l

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