ML20059N016

Proposed TS 4.4.5.0 Reflecting Footnote Which Addresses 931024,Unit 1 Unplanned Outage Which Was Needed to Identify & Repair Tube Leak on SG 1C

ML20059N016
Person / Time
Site:	Braidwood
Issue date:	11/12/1993
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20059N009	List: ML20059N007 ML20059N016
References
NUDOCS 9311300058
Download: ML20059N016 (9)

Text

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2: . ' ATTACHMENT B PROPOSED CHANGES TO APPENDIX A' FOR FACILITY OPERATING LICENSES NPF-72 and NPF-77 Li i

.i Revised Page: -t 3/4 4-13 ,

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93113'00058 DR 93111p"*

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y REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLEl - -

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable steam i generator (s) to OPERABLE status prior to increasing T,yg above 200 4.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Speci fication 4. 0.5.44- - - - . - -

.l 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be ,ceterminec OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tuoe minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall' include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry

' indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

b. The first sample of tubes selected for ea:h inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

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l l BRAIDWOOD - UNITS 1 & 2 3/4 4-13 E F E G LW CT W . l

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1. Unit 1 entered an unplanned outage (A1F26) on October 24,1993, to repair a tube leak in the 1C Steam Generator. The tube leak was less than the reactor-to-secondary leakage limit of Specification 3.4.6.2c for one steam generator.

The generator was determined to be OPERABLE following completion of the inspection plan detailed in Letter # SVP/93-063, S. M. Berg, Jr. (CECO) to J.

Zwolinski (NRC), dated November 10,1993. The generator shall be demonstrated OPERABLE in accordance with Specification 4.4.5.0 prior to the initial resumptiom of plant operation following the Unit 1 Cycle 4 Refuel Outage (A1 RO4).

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. ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Commonwealth Edison Company (CECO) has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10, Code of Federal Regulations, Part 50, Paragraph 92, Subparagraph c, [10 CFR 50.92(c)], a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

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

The proposed amendment would add a footnote to TSSR 4.4.5.0 to reference the inspection program used to demonstrate the OPERABILITY of the 1C Steam Generator following the unplanned outage (A1F26) which began October 24,1993, to repair a tube leak which was less than the reactor-to-secondary leakage limit of Specification 3.4.6.2c for one steam generator. The result of this inspection program will satisfy OPERABILITY requirements until the next scheduled steam generator tube inservice inspection to be performed during the Unit 1 Cycle 4 Refuel Outage (A1RO4) currently scheduled to begin March 5,1993.

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Following identification of the flaw in Tube 49/76 (Row 49, Column 76) of the 1C Steam Generator, a comprehensive and conservative inspection program was developed. Tais program included the following actions:

a. conduct rotating pancake coil (RPC) eddy current testing of the tubes surrounding Tube 49/76 between the third anti-vibration bar (AVB) and the top support plate in the cold leg. This action was necessary to resolve the possibility of physical damage in the vicinity of the leak,

b. perform a 100% fulllength bobbin coil eddy current inspection on all

. tubes in the 1C Steam Generator. This was performed to identify other freespan indications that may be precursors to the flaw observed in Tube 49/76,

c. if other freespan indications similar in nature to the failed tube were identified, a new degradation mechanism could be indicated, and consideration would be given to inspections in the other three steam generators,

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4 9' . ATTACHMENT C y 3

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS

d. If no freespan indications were identified, then the failure of Tube 49/76 would be considered an isolated event, e, it was fully expected to find other eddy current indications (particularly at. 3 the support plates on the hot leg side) which under planned i inservice inspection (ISI) conditions would be dispositioned i by Commonwealth Edison Company (CECO) guidelines. .

For the purposes of the current situation, the Technical a Review Team decided that the appropriate and prudent response to these indications would consist of:

1) plug all clear indications of greater than or e. qual to 40% through- 'i wall degradation consistent with the Technical Specification Plugging Limit of Specification 4.4.5.4, and {

2) plug any distorted indications which showed abnormal growth a when compared to data from previous outages. l The RPC eddy current testing of the tubes in the vicinity of the leak showed no indication similar in nature to flaw. ,

The bobbin coil eddy current testing of 100% of the available tubes in the 1C' 4

Steam Generator identified 17 freespan indications. Each^ indication was furtherj evaluated by RPC eddy current testing. It was determined that none of these indications were similar in nature to the flaw in Tube 49/76. Based on the resultc of these inspections, it was determined that the flaw in Tube 49/76 was an isolated incident. There was no reason to believe that'a flaw similar in' nature to that in Tube.49/76 would be identified if the other steam generators were inspected.

Historically, the 1C Steam Generator has exhibited the most tube degradation,: ,

when compared to the other three steam generators (See Table .1). Since the : [

results of the inspection conducted during this unplanned outage were  ;

consistent with the expected degradation since the last steam generator tube ,

inservice inspection, it was decided that it'was not necessary to accelerate the'-  ;

, scheduled steam generator tube inservice inspections for the other steam gen _erators at this time to monito'r tube degradation. Th's last steam generator-tube inservice inspections, performed during A1RO3,'still provided sufficienti i assurance that Unit 1 steam generators could be safely. operated until the next scheduled steam generator tube inservice inspection scheduled to be performed during A1RO4.

- The~ issuance of this proposed amendment will have no impact on the" consequences of any accident analyses.

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ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS As a result of this event, the following compesatory actions have been implemented:

a. steam jet air ejector and main steam line radiation monitors alert and alarm setpoints have been lowered, Old New Main Steam Line alert .25mR/hr .2mR/hr alarm .5mR/hr .4mR/hr SJAE* alert 5.75E 6 3 E'6 alarm 1.15E~5 1 E'S

• Micro Curies /cc

b. the abnormal operating procedure for primary to secondary leakage has ,

been revised to include the following:

1) guidance to operations and chemistry personnel regarding monitoring to detect increases in leakage rate when known primary to secondary leakage exists. This guidance includes the use of the portable N-16 monitor, and

2) an administrative limit for primary-to-secondary leakage greatar than 150 gpd or a change in leak rate of greater than 25 gpd in 1 >

hour would require a plant shutdown. This limit is below the Technical Specification limit and was chosen based upon the best technical information available, and

c. operating shifts will be briefed on the requirements of the revised abnormal operating procedure prior to assuming duties when the Unit 1 reactor is critical.

These actions will enhance the plant staff's ability to detect, monitor, and appropriately respond to small steam generator tube leaks.

The steam generator tube rupture analysis will continue to be bounding in this case.

Therefore, the proposed amendment does not significantly increase the probability _ or consequences of an accident previously evaluated.

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_ 1 T ". . ATTACHMENT C -

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS

2. The proposed change does not create the possibility of a new or different kind - .;

of accident from any accident previously, implementation of the proposed amendment will not introduce any significant or adverse changes to the plant design basis. Any accident as a result of.

potential tube degradation is bounded by the existing steam generator tube .

rupture analysis. Therefore, implementation of the proposed amendment does not create a new or different kind of accident.  :

3. The proposed change does not involve a significant reduction in a margin of j safety. 1 The inspection program that was implemented as a result of this event' indicated

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with a high degree of certainty that the failed tube was an isolated event. This  :

result was based upon rotating pancake coil eddy current testing that was performed for each freespan indication. No indication similar in nature to the  :

flaw was identified. 1 An appropriate portion of CECO's inservice inspection Program wac performed.

This inspection program applied Braidwood's Technical' Specification criteria for-- ,

the plugging limit, as outlined in Specification 4.4.5.4. Additionally, the ,

inspection program included plugging any distorted indications which showed abnormal growth when compared to data from previous outages. Ttie results of , j this inspection were consistent with what was expected.

a Historically,~ the 1C Steam Generator has exhibited the most tube degradation,- j when compared to the other three steam generators (See Table 1). Since the o results of the inspection conducted during this unplanned outage were consistent with th.e' expected degradation since the last steam generator tube.

Inservice inspection,-it.was decided that it'was not necessary to accelerate the; +

scheduled steam generator tube inservice inspections for the other. steam ,

generators at this time to monitor tube' degradation'. The last' steam generator - j tube inservice inspections, performed during A1RO3, still provided sufficient- ~i assurance that Unit 1 steam generators could be safely operated until the~ next- ,

sche.duled steam generator tube inservice inspection scheduled to be' performed during A1RO4. .c Therefore, the proposed amendment wii! not involve a significant reduction in a margin of safety.

Based upon the above evaluation, CECO has concluded that this propcsod amendment involves no significant hazards considerations. -

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'. ATTACHMENT D ENVIRONMENTAL ASSESSMENT- ,

Commonwealth Edison Company has evaluated the proposed amendment against the criteria for the identification of licensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Fedreal Regulations, Part 51, Paragraph 21 (10 CFR 51.21). It has been determined that the proposed change meets the criteria for a categorical exclusion as provided for under 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, and that the change requested involves changes to surveillance requirements, and involves no significant hazards considerations as discussed in Attachment C. There is no change in the types, or significant increase in the amount, of any effluents that may be released offsite. There is no significant increase in individual or cumulative occupational radiation exposure.

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TABLE.1 BRAIDWOOD UNIT 1 PLUGGING HISTORY t

STEAM STEAM STEAM STEAM.

GEN "A" GEN "B" GEN "C" GEN "D" FACTORY / 4 1 8 1 PSI

• A1 RO1 5 1 0 0 A1 R02 11 2 19 4 ,

4-A1R03 37 11 82 44.

10/93 OUTAGE N/A N/A 117 .N/A 3

PRE-SERVICE INSPECTION .

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