ML20079L537

Proposed Tech Specs Re Reduction of primary-to-secondary Leakage Limit to 140 Gallons Per Day Through SG or 420 Gallons Per Day Through All Three SGs

ML20079L537
Person / Time
Site:	Farley
Issue date:	10/29/1991
From:	ALABAMA POWER CO.
To:
Shared Package
ML20079L534	List: ML20079L528 ML20079L537
References
NUDOCS 9111060371
Download: ML20079L537 (7)

Text

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Proposed Changed Technical Specification Pages Vall 1 Page 3/4 4-17 Replace Page D 3/4 4-3 Replace Page B 3/4 4-4 Replace 91110/.0371 911029 DR ADOCK0500(({j,9

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• RfAC10R COOLANT SYSl@

0ELR2110ML LEAKA_K LIM 111NG COND1110N FOR OPERA 110N 3.4.7.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM VNIDENilfiED LEAKAGE,

c. 420 gallons per day total primary-to-secondary leakage through all steam generators and 140 gallons per day through any one steam generator,

d. 10 GPM 10ENilflED LEAKAGL from the Reactor C0olant System, and

e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig.

f. The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 223S 1 20 psig.

APPLICABILITYL MODES 1, 2, 3 and 4 ACTION: a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHU100WN 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 leakage 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 1101 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 SHU100WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System Pressure isolation Valve leakage greater than the limit specified in Table 3.4-1, isoute 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 closed manual or deactivated automatic valves, 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 SHUIDOWN 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 4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atinosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Monitoring the containment air cooler condensate level system or containment atmosphere gaseous radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

i FARLEY - UNIT 1 I 3/4 4-17 AMENDMENT N0.

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• REAC10R CQQWILjiyJi][tj .

'BASLS 3/4.4.6 $1EAM LUILPRORS ,

lho Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or Inservice conditions that lead to 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes, if the secondary coolant chemistry is not maintained within these limits, localized terrcsion may likely result in stress corrosion cracking. 1he extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage - 140 gallons per day per l steam generator). Cracks hav'ng a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operational leakage of this magnitude can be readily detected by existing farley Unit I radiation monitors. Leakago in excess of this limit will require plant shutdown and an unschedulcd inspection, during which the leaking tubes will be located and plugged or repaired.

Whstage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during sc' lied inservice steam generator tube examinations.

plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. If a s1 coved tube is l found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 37% for the laser welded sleeve of sloove nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R.G.1.121 calculations with 20% added for conservatism.

The portion of the tube and the sloove for which indications of wall degradation must be evaluated can be summarized as-follows;

a. Mechanical-

1. Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.

2. Indication of tube degradation of any type including a complete guillotine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.

FARLEY-UNil 1 B 3/4 4-3 AMENDMENT NO.

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• IDC10R _CQQL6RL1Hld

' BASES

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3 / 4 . 4 . 7 R E &C10lL[QOlafLLS111M_LlIM(

ULIAL1[ALMLDf1LCl10tLHSlW2 1he RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the ret.ommendations of Regulatory Guide 1.45,

• Reactor Coolant Pressure Boundary Leakage Detection Systems, Hay 1973.

UsALLQPERA110NAL LEAMGL Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPH. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTiflED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDEN11Fl[D LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactw coolant pump seals exceeds 31 GPM with the modulating valve in the supply lind fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The survelliance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity, thereby reducing the arobability of gross valve failure and censequent intersystem LOCA. Lea (age from the RCS Pressure Isolation valves is IDENTIFl[D LEAKAGE and will be consideret a portion of the allowed limit.

The total steam generator tube leakage limit of 420 gallons per day l for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. A1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 140 gallons per day leakage limit per steam generator l ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

PRESSURE B0VHDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SilVTDOWN.

FARLEY-UNil 1 B 3/4 4-4 AMENDMENT NO.

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Attachment 2 Significant Hazards Evaluation 4

f arley Neclear Power Plant Unit 1 Reduction in the Ste*w Generator Primary-to-Secondary leakage Limit Significant Hazards Evaluation IN1RODUCT10N AND BACKGROUND inspections of the steam generators at f arley Nuclear Plant (INP) Unit I have revealed the presence of circumferential1y oriented indications in the WEX1[X expansion transition in tubes within the sludge pile region, and the U-bend region of row I tubes in steam generators A, 8, and C. As a result of the circumferential cracks, a conservative 140 gallons per day leakage rate limit was established for Unit 1. Currently, Technical Specification 3.4.7.2.c allows up to 500 gallons per day through any one steam generator.

DESCRIPil0N Of THE AMENDMENT R[ QUEST As required by 10 CFR 50.91(a)(1), an analysis is provided to demonstrate that the proposed license amendment to reduce the steam generator primary-to-secondary leakage limit for fariey Unit 1 involves no significant hazards. The current technical specification allows 1 gallon per minute total primary-to- .

secondary leakage through all steam generators and 500 gallons per day through any one steam generater. By letter dated February 26, 1991, Alabama Power Company requested a technical specification amendment to utilize steam generator tube support plate alternate plugging criteria. 1his February 26, 1991 amendment request proposed reducing the leakage limits to 450 gallons per day ,

total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator. This amendment request proposes to further reduce the leakage limit to 420 gallons per day total primary-to-secondary leakage through all steam generators and 140 gallons per day through any one steam generator.

ANALYSIS in accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amendment is analyzed using the following standards and found not to:

1. involve a significant increase in the probability or consequences for an accident previously evaluated; or

2. create the possibility of a new or different kind of accident from any accident previoLsly evaluated; or

3. involve a significant reduction in a margin of safety.

Conformance of the proposed amendment to the standards for determination of no significant hazards as defined in 10 CFR 50.92 is demonstrated below. ,

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1. Operation of farley Unit 1 in accordance with the proposed Itcense amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The license amendment reduces the primary-to-secondary leakage limit for steam generators. No physical changes will be made to the plant. A reduction in the leakage limit will result in more conservative operation of the plant requiring an eariter shutdown for steam generator leakage.

As a result, neither the probability or consequences of any previously evaluated accident will be increased.

2. The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Implementation of the reduced primary-to-secondary leakage limit wi'll not introduce any physical changes to the plant. Use of the reduced leakage limit will result in a more conservative response to primary-to secondary steam generator leakage.

3. The proposed Itcense amendment does not involve a significant reduction in margin of safety.

The use of the reduced primary-to-secondary steam generator leakage limit will result in improved margin to steam generator tube failure. Reducing i the allowed leakage limit to 140 gallons per day will result in more conservative o)eration of the plant since unit shutdown will be required at a lower lea cage level.

CONCLUSION Based on the preceding analysis, it is concluded that the reduction in the steam generator primary-to-secondary leakage limit to 140 gallons per day for farley Unit 1 is acceptable and the proposed license amendment does not involve a Significant Hazardo Consideration finding as defined in 10 CFR 50.92.

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